'EPRI PWR Safety & Relief Valve Test Program:Test ... · NOTICE This report was prepared by the Electric Power Research Institute, Ineo (EPRI). Neither EPRI, members of EPRl, nor

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EPRI PWR SAFETY AND RELIEF VALVE TEST PROGRAM

TEST CONDITION JUSTIFICATION REPORT

Research Project Vl02 Interim Report, April 1982

Prepared by

PWR VALVE PROGRAM STAFF Nuclear Power Division

Principal Investigator J. Hosler

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8207160355 820401 PDR ADOCf.( 05000255 P PDR

Prepared for

PARTICIPATING PWR UTILITIES

and

ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue

Palo Alto, California 94304

EPRI Project Manager J. Hosler

Nuclear Power Division

NOTICE

This report was prepared by the Electric Power Research Institute, Ineo (EPRI). Neither EPRI, members of EPRl, nor any person acting on their behalf: (a) makes any warranty, ex~ress or illll)lfed, with respect to the use of any 1nfonnationw apoaratusi method, or orocess disclosed in this report or that such use may not infr1nqe DF1vate1y owned rights; or (b) assunes any 11ab11it1es with respect to the use of v or for damages resulting from the use ofp any information, apparatus, method, or process disclosed in this reporto

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ABSTRACT

In response to NUREG 0737, Item II.O.l.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented •

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FORWARD

This report was prepared by the EPRI PWR Valve Program Staff to assist utilities participating in the EPRI PWR Safety and Relief Valve Test Program in justifying that the conditions under which their safety and relief valve designs (or valve designs representinq their valves) were tested are representative of those expected in their units.

The report references the "Plant Conditions Justification Reports" prepared under EPRI contract by the PWR NSSS vendors as well as the EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report.

The information in this report, together with that contained in the Plant Conditions Justification, Valve Selection/Justification, and Safety and Relief Valve Test Reports generated under the aforementioned program, constitutes a sufficient basis to judqe the operability of all the primary system safety and relief designs utilized by participating PWR utilities under the full range of expected fluid inlet conditions.

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CONTENTS

Section Title

1.0 INTRODUCTION

2.0

3.0

1.1 Background 1.2

1.3

Objective Scope

TEST CONDITION SELECTION APPROACH 2.1 General 2.2 Relief Valves

? ? 1 Fluid Chemistry ....... 2.2.2 Inlet Piping 2.2.3 Water Seals 2.2.4 Back Pressure -2.2.s Valve Loading

2.3 Safety Valves 2.3.l Fluid Chemistry 2.3.2 Inlet Piping 2.3.3 Loop Seals 2.3.4 Back Pressure 2.3.5 Valve Loading 2.3.6 Liquid Surge Flow 2.3.7 Ring Positions

JUSTIFICATION OF RELIEF VALVE TEST CONDITIONS 3 .1 General 3.2 Dresser

3.2.l* FSAR Events 3.2.2* Extended High-Pressure Injection Events 3.2.3* Cold Overpressurization Events

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Page

1-1 1-1 1-1 1-2

2-1 2-1 2-2

. 2-2

2-3 2-3 2-3 2-4 2-4 2-4 2-5 2-6 2-6 2-6 2-6

3-1 3-3

Section Title Page •

3.3 Crosby 3-11 •• 3.4 Target Rock 3-18 3.5 Control Components 3-25

3.6 Masonef lan J..33

3.7 Copes .. vulc:an U7-14PH Plug and Cage) 3=42

3.8 Copes-Vulcan (316w/Ste111te Plug and l7~4PH Cage) 3.,,49

3.9 Muesco Controls 3~57

3.10 Fishel'" Controls 3-63 3.11 Garrett 3-71

4.0 JUSTIFICATION OF SAFETY VALVE TEST CONDITIONS 4-1 4.1 General 4-1 4.2 Dresser 31709NA 4=6

4.2.1* FSAR Events 4.2.2* Extended High Pressure Injection Events

4.3 Dressel'" 31739A 4 .. 10

4.4 Crosby 3K6 (Steam Internals) 4~15

4.5 Crosby 3K6 (Loop Seal Internals) 4-19 4.6 Crosby 6M6 (Steam Internals) ~23

4.7 Crosby 6M6 (Loop Seal Internals) 4-24 4.8 Crosby 6N8 (Steam Internals) 4-=28

4.9 Crosby 6N8 (Loop Seal Internals) 4 .. 32 4.10 Target Rock 69C 4-33

s.o REFERENCES 5~1

*Same subsection titles repeated for each valve type.

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Table

1-1 3-1 3=2

3-3

3-4 3-5 3-6

3-7

3-8 3-9

3-10

3-11 3-12

3-13

3-14 3-15

3-16

3-17

3-18

3-19

3-20

TABLES

PWR Units Considered in Determination of Valve Test Conditions "As Tested" Marshall Test Matrix for the Dresser Relief Valve "As Tested" Wyl e Phase II Test Matrix for the Dresser Relief Valve "As Tested" Wyle Phase III Test Matrix for the Dresser Relief Valve PWR Units with Dresser Relief Valves "As Tested" Marshall Test Matrix for the Crosby Relief Valve "As Tested" Wyle Phase II Test Matrix for the Crosby Relief Valve "As Tested" Wyle Phase III Test Matrix for the Crosby Relief Valve PWR Units with Crosby Relief Valves "As Tested" Marshall Test Matrix for the Target Rock Relief Valve "As Tested" Wyle Phase III Text Matrix for the Target Rock Relief Valve PWR Units with Target Rock Relief Valves "As Tested" Marshall Test Matrix for the Control Components Relief Valve "As Tested" Wyl e Phase II I Test Matrix for the Control Components Relief Valve PWR Units with Control Components Relief Valves "As Tested" Marshall Test Matrix for the Masoneilan Relief Valve "As Tested" Wyle Phase III Test Matrix for the Masoneilan Relief Valve PWR Units with Masoneilan Relief Valves "As Tested" Marshall Test Matrix for the Copes-Vulcan (17-4PH Plug and Cage} Relief Valve "As Tested" Wyle Phase III Test Matrix for the Copes-Vulcan {17-4PH Plug and Cage} Relief Valve PWR Units with Copes-Vulcan (17-4PH Plug and Cage) Relief Valves

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Page

1-3 3-7

3-8

3-9

3-10 3-14

3-15

3-16 3-17

3-22

3-23 3-24

3-29

3-30

3-32

3-38

3-39 3-41

3-46

3-47

3-48

Table Page

3-21 "As Tested" Marshall Test Matrix for the Copes-Vulcan .., (316 w/stellite Plug and 17-4PH Cage) Relief Valve 3-54

3-22 "As Tested" Wyle Phase III Test Matrix for the Copes-Vulcan (316 w/stellite Plug and 17-4 PH Cage) Relief Valve 3-55

3-23 PWR Units with Copes-Vulcan (316 w/stellite Plug and 17-4PH cage) Relief Valves 3-56

3-24 •As Tested• Marshall Test Matrix for the Muesco Relief Valve 3-60 3-25 "As Tested" Wyle Phase III Test Matrix for the Muesco

Controls- Relief Valve 3-61 3-26 PWR Units with l'f.lesco Controls. Relief Valves 3-62 3-27 "As Tested" Marshall Test Matrix for the Fisher Controls

Relief Valve 3-68 3-28 "As Tested" Wyle Phase III Test Matrix for the Fisher

Controls Relief Valve 3-69 3-29 PWR Units with Fisher Controls Relief Valves 3 ... 70 3-30 •As Tested• Marshall Test Matrix for the Garrett Relief Valve 3-76 3-31 "As Tested• Wyle Phase III Test Matrix for the Garrett Relief

Valve 3-77 3-32 PWR Units with Garrett Relief Valves 3-78

1··-· .. 4-1 Limiting Range of Safety Valve Inlet Fluid Conditions for PWR Units Listed in Table 1-1 4-4

.: 4-2 Sunmary of Safety Valve Inlet Piping Configurations Tested 4-5 4-3 "As Tested" Combustion Engineering Matrix of Tests Perfonned

on the Dresser 31709NA Safety Valve with ~Reference" Ring Positions 4-9

4-4 "As Tested• Combustion Engineering Matrix of Tests Performed on the Dresser 31739A Safety Valve with "Reference" Ring Positions 4-13

4-5 .,As Tested" Combustion Engineering Matrix of Tests Perfonned on the Crosby 3K6 (Steam Internals) Safety Valve with •Reference 11 Ring Positions 4~18'

4-6 "As Tested• Combustion Engineering Matrb of Tests Performed . . on the Cr.osb,y .3!<6 (loop Seal Internals)' Safety Valve with ~Reference'' Ring Positions 4-22

4-7 •As TestedA Combustion Engineering Matrix of Tests PeFformed on the Crosby 6M6 (Loop Seal Internals) Safety Valve with "Reference" Ring Positions 4-27

4-8 "As Testedm Combustion Engineering Matrix of Tests Performed an the Crosby 6N8 (Steam Internals) Safety Valve with .,Reference• Ring Positions 4=31

4-9 11 As Tested" Combustion Engineering Test Matrix for the Target Rock 69C Safety Valve 4--36

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SUMMARY

The Nuclear Regulatory Commission (NRC), in NUREG 0578 and NUREG 0737, required utilities operating and in the process of constructing pressurized water reac-tor (PWR) power plants to perfonn a test program to demonstrate the operability of pressurizer power operated relief valves (PORVs) and safety valves. In response to these requirements, the PWR utilities assigyed the Electric.Power Research Institute (EPRI) the responsibility for conducting a comprehensive test program to obtain full-scale operability data, applicable to all PWR safety and relief valve designs, under the full range of conditions under which they may be expected to operate.

As part of their response to the NUREG requirements, each PWR utility was required to provide evidence that the inlet fluid conditions under which each of thei~ valve designs was tested as part of the EPRI program are representative of those expected in their units. In addition, the testing was to address discharge piping effects on valve operability.

To detennine the valve inlet fluid conditions to be selected for testing, EPRI contracted with each PWR NSSS vendor to provide (via "Plant Conditions Justification Reports") the expected range of safety and relief valve inlet fluid conditions for FSAR, Extended High Pressure Liquid Injection, and Cold Overpressurization events in units of their design.

The objective of this report is to provide justification that the inlet fluid conditions under which each of the safety and relief valve designs was tested are representative of those expected for the events named above in participating PWR units and to discuss the discharge piping effects on valve operability addressed in the perfonnance of the test program. It should be noted that the inlet fluid conditions under which testing was perfonned are .representative of the most limiting conditions expected in all participating PWRS. Therefore, some of the conditions under which testing was perfonned are not applicable to some PWR units.

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It is expected that this report, as well as the "Plant Condition Justification Ret1orts 11 generated by the PWR NSSS vendors, will be referenced by participating PWR utilities in the portion of their plant-specific submittals justifying the applicability of the conditions tested to their units.

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• Section 1.0

INTRODUCTION

1.1 BACKGROUND /

The Nuclear Regulatory Commission (NRC), in NUREG 0578 and NUREG 0737, required utilities operating and in the process of constructing pressurized water reactor (PWR) power pl ants to perform a test program to demonstrate the operabi 1 ity of· pressurizer power operated relief valves (PORVs) and safety valves. In response to these requirements, the PWR utilities assigned the Electric Power Research Institute (EPRI) the responsibility for conducting a comprehensive test program to obtain full-scale operability data, applicable to all PWR safety and relief valve designs, under the full range of conditions under which they may be expected to operate.

As part of their response to the NUREG requirements, each PWR utility was required to provide evidence that the inlet·fluid conditions under which each of their valve. designs was tested are representative of those expected in their units. In addition, the testing was to address discharge piping effects on valve operability.

To determine the valve inlet fluid conditions to be selected for testing, EPRI contracted with each PWR NSSS vendor to provide (via "Plant Conditions Justification Reports") the expected range of safety and relief valve inlet fluid conditions for FSAR, Extended High Pressure Liquid Injection, arid Cold Overpressurization events in ~nits of th~ir d~sign.

1.2 OBJECTIVE

The objective of this report is to provide justification that the inlet fluid conditions under which each of the safety and relief valve designs was tested are representative of those expected for the events named above in participating PWR units and to discuss the discharge piping effects addressed in the performance of the test program. It should be noted that the inlet fluid conditions under which testing was performed are representative of the most limiting conditions expected in all participating PWR units. Therefore some of the conditions under which

• testing was performed are not applicable to some PWR units.

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It is expected that this report, as well as the 11 Pl ant Condition Justification Reports" generated by the PWR NSSS vendors, will be referenced by participating PWR utilities in the portion of their pl ant•specific submittals just.ifying the applicability of the conditions tested to their units.

lc3 SCOPE

This report provides justification that the inlet fluid conditions under which ea~h valve design was tested, as part of the EPRI/PWR Safety and Relief Valve Test Program, are representative of those expected for FSAR, Extended High Pressure Liquid Injection and Cold Overpressurization events in all B&W and most Combustion Engineering and Westinghouse nuclear units. Table 1-1 lists the PWR units considered in detennination of the conditions to be testedc It should be noted that the "Plant Condition Justification Reports" prepared by the PWR NSSS vendors did not provide expected conditions resulting from Cold Overpressur1zation events fo~ some of these units either because the vendor did not design the system or because the plant specific analysis had not been completed at the time the reports were prepared (see note 1 Table lal)& For these units, justification of the applic cability of EPRI test conditions to those expected for such events will be provided as part of their plant specific evaluationso

The test values for other parameters that may affect the operability of relief and safety valves~ i.e., inlet piping geometry (safety valves only), and discharge piping {back pressure and induced bending moment) were selected to be representam tive of those existing 1n PWRs. Justification of the applicability of the tested

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values for these parameters to specific in-plant installations will be included as part of each participating PWR utility• s·. ·pl ant~specific: evaluation.,.

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Table 1-1

PWR UNITS CONSIDERED IN DETERMINATION OF VALVE TEST C.ONDITIONS

UNIT

Oconee 1, 2, 3

Three Mile Island 1, 2

Crystal River 3

Rancho Seco

Arkansas Nuclear One - 1

Davis-Besse

Midland 1, 2

Bel 1 efonte 1, 2

Washington Nuclear Project 1, 4


Calvert Cliffs 1, 2

Maine Yankee*

Millstone Point 2

Fort Ca 1 houn

Palisades

St. Lucie 1, 2

San Onofre 2, 3

Waterford 3

Palo Verde 1, 2, 3

Yellow Creek 1, 2

Washington Nuclear Project 3, 5

Cherokee 1, 2

Perkins 1, 2 and 3

NSSS VENDOR

Babcock and Wilcox

Combustion Engineering

*No justification for Cold Overpressurization Conditions provided herein. Such justification will be provided as part of this Utility's plant specific evaluation.

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Table 1-1 (continued)

PWR UNITS CONSIDERED IN DETERMINATION OF VALVE TEST CONDITIONS

UNIT

R. E. Gf nna*

Pofnt Beach* 1, 2

Prairie Island* 1, 2

Kewaunee*

San Onofre l*

H. B. Robinson*

Turkey Point 4*

SUM'Y* 1, 2 .

Beaver Valley* 1, 2

North Anna* 1, 2

Joseph M. Farley* l, 2

Virgil C. Sunmer l*

Shearon Harris* 1, 2, 3, 4

Indian Point* 2, 3

Zionw 1, 2

Donald c. Cook* 1,2

D1ablo Canyon* l, 2

Trojan*

Sequoyah 1, 2

Salem* 1, 2

NSSS VENDOR

Westinghouse

*No justification for Cold Overpressurization Conditions provided herein. Such justification will be provided as part of these Utilities' plant specific evaluations. '

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• Table 1-1 (continued)

PWR UNITS CONSIDERED IN DETERMINATION OF VALVE TEST CONDITIONS

UNIT

W. B. McGuire* 1, 2

Byron 1, 2

Alvin W. Vogtle* 1, 2

Mi 11 stone 3*

Seabrook* 1, 2

Catawba* 1, 2

Braidwood 1, 2

South Texas 1, 2

Marble Hill* 1, 2

Wolf Creek

Call away 1, 2

Comanche Peak 1, 2

Watts Bar 1, 2

NSSS VENDOR

Westin house

*No justification for Cold Overpressurization Conditions provided herein. Such justification will be provided as part of these Utilities' plant specific evaluations.

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Section 2.0

TEST CONDITION SELECTION APPROACH

2.1 GENERAL

The approach used in develo(lnent of inlet fluid conditions for relief valve testing was to select, for each specific valve design tested, a set of conditions which would be representative of those expected in all participating PWR units currently utilizing that valve design or one represented by it. A similar approach was used in dev.elo(lnent of safety valve test conditions except that a single set of limiting fluid inlet conditions was selected which envelops those expected in all of the PWR .units listed in Table 1-1.

The sources of infonnation used to develop the range of test conditions included the PWR NSSS vendor "Pl ant Condition Justification Reports (References 1, 2, and 3), valve type and piping configuration data provided by participating PWR Utilities, and scoping assessments of expected valve back pressures.

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2.2 RELIEF VALVES

The following parameters are considered to have a potential for affecting the operability of power operated relief valves:

o Conditions existing when Valve is Opening/Closing.

Inlet Fluid State/Pressure/Temperature.

Bending Moment Induced in.the Valve Body.

o Back Pressure Developed at the Valve Outlet.

o Fluid Chemistry

Al 1 of the above except fluid chemistry were selected as primary test parameters. Test values for inlet fluid state, pressure, temperature were selected to represent those expected in participating PWRs. Test values for the other primary test parameters are.considered to be representative of those in PWRs; however, as discussed in Section 1.3, specific justification that such conditions apply to specific PWR units wil 1 be included in each participating uti1 ity 1 s pl ant specific evaluation. The following sections expand on the approach used to define re.lief valve test parameters.

2.2.1 FLUID CHEMISTRY

Evaluation of the effect of fluid chemistry on valve operability.is not considered to be within the scope of this short-term valve operability demonstration program. Such effects (if any) are considered to be lon~term and could only be evaluated as part of an extended valve aging study.

2.2.2 INLET PIPING

PORVs used in PWRs are signaled to open by a pressure sensed in the main volume of the pressurizer or in the hot leg. As a result, any transient pressure fluctuations at the inlets of such valves due to the characteristics of the valve inlet piping, would not affect valve operability. Therefore, the relief valve inlet piping geometries tested need not be directly representative of those found in participating PWRs.

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2.2.3 WATER SEALS

The primary effect of a water seal on relief valve operation is considered to be the thermal shock induced as the cold water is followed by hot steam through the· valve internals. To simulate this effect, cold (-100°F) to hot (-650°F) water transition tests were perforined on each relief valve design, where applicable.

2.2e4 BACK PRESSURE

Back pressure developed at the outlet of PORVs is considered of potential importance to valve operability only for the pilot operated relief valve designs, i.e., Dresser, Crosby, Target Rock, and Garrett.

The air operated PORVs.are not considered to be sensitive to back pressure since they are designed to fully open and close with zero delta pressure between the inlet and outlet of ·the valve which is equivalent to developing a back pressure equal to the inlet pressure. For these valves, justification of the typicality of the back pressures tested is not required in p1ant00specific evaluations.,

2e2.5 VALVE LOADING

The valve loading considered to have the greatest potential for affecting relief valve operation is bending moment induced across the valve body while the valve is opening or closinge Such loading could potentially result in mechanical binding of internal moving parts. Each relief valve design was tested at least once under an exte~na11y applied static bending moment. In addition, during such tests, a

-·- dynamic bending moment was induced by the system which combined with the static moment to provide loads generally in the range of ASME code allowables for primary bending for the corresponding attached piping sizes and materials utilized in PWRs.

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2 .. 3 SAFETY VALVES The following parameters are considered to have a potential for affecting the operability of safety valves:

o Inlet Fluid Conditions While Valve is Opening/Closing Fluid State/Pressure/Temperature

-- Driving Liquid F1owrate (if valve is passing liquid) -- Bending Moment Induced in the Valve Body

o Peak Pressure Attained at the Valve Inlet o Backpressure Developed at the Valve Outlet o Inlet Piping Geometry o Valve· Ring Positions o Fluid Chemistry

All of the above were selected as primary test parameters except Fluid Chemistry.

The test values for inlet fluid state, pressure and temperature were strictly controlled to represent those expected in participating PWR units. Test values for the other primary test parameters were in the range of those expected in PWRs. The following sections expand on the approach used to select safety valve test parameters and their values.

2.3.l FLUI"O CHEMISTRY

As explained in section 2.2.l, evaluation of the effects of fluid chemistry on valve operability_ was not considered to be within the scope of this short-term valve operability demonstration program.

2.3.2 INLET PIPING

Safety valve inlet piping geometry (length and diameter of piping) has the potential for affecting safety valve operation. Inlet piping that is long or of a small diameter may cause a transient drop on valve opening in the total pressure at the valve inlet approaching or exceeding the blowdown for that valve. On valve opening, such a drop in inlet pressure would introduce the potential for an immediate reclosure of the valve. Upon valve closure, a fluid hammer at the valve inlet would occur which could result in a reopening of the valve. This cycle may repeat at high frequency, a phenomenon known as "chatter".

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Because a range of safety valve inlet geometries exists in PWRs and because the inlet piping geometries tested included area changes not generally found in PWRs (due to flow venturi installation), a method was required to assess the applicability of test results obtained with the inlet piping geometry/safety valve design combinations tested, to those existing in PWRs.

PRESSURE DROP AT VALVE INLET

The method developed consists of a comparison between the expected total pressure drop for the in-plant safety valve/inlet piping combination to that for the safety valve/inlet piping combination tested. The total pressure drop fs comprised of a frictional and an acoustic wave component and is evaluated under steam flow conditions.

USE OF COMPUTED TOTAL PRESSURE DROP

The computed fn-plant total pressure drop is compared to that for the test eonfi~ guration. If the in-plant pressure drop is less than the corresponding value computed for a test configuration (with the same valve design or one represented by it) 1 the in-plant valve performance should be at least as stable as that of the test valve.

2.3.3 LOOP SEAL

Many safety valve installations 1n PWRs incorporate a '1loop seal 11 between the pressurizer and the valve inlet. This piping arrangement results in the collection of relatively cold water at the inlet of the safety valve. The test program included tests, where applicable, with loop seal piping arrangements and loop seal water temperatures representative of those found in PWR units.

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2.3.4 BACK PRESSURE

The back pressure developed at the outlet of safety valves is considered a primary parameter affecting its performance. Higher back pressures would have a greater tendency to affect valve lift and reclosure characteristics. Therefore, the back pressures selected for test included values at or above those expected. in PWRs based on scoping evaluations •

. 2.3.5 VALVE LOADING

·During the performance of the safety valve tests, substantial bending moments were induced in the valve body due to thermal expansion of attached discharge piping. These moments were greatest at the time of valve closure and can be used to demonstrate the valve's ability to operate under an imposed bending moment.

2.3.6 LIQUID SURGE FLOW

The liquid surge rates driving the opening of safety valves in PWRs are less than the full open choked flow capacity of these valves: Therefore, the actual flow through these valves will be determined by the intermediate degree of opening of the valve which is considered to be primarily a function of inlet pressure. Since inlet pressures representative of those expected in PWRs were tested, strict control of the driving liquid surge flow was not deemed necessary, although surge flows in the range of those expected were tested.

2.3.7 RING POSITIONS

. Each spring loaded safety valve design tested has two or three (depending on the valve manufacturer) notched rings which, when adjusted, affect the geometry of the "huddling chamber" within the valve. The position of these rings affects the opening time, lift and blo\'itlown of the valve •. As discussed in Section 2.3.2, th.e blowdown may affect valve stability.

In general, each spring-loaded safety valve was initially tested with ring positions based on manufacturer reconmendations and representative of those utilized in typical plant installations. In addition, several spring loaded safety valves were tested over a range of ring positions to try to improve performance under steam flow-conditions.

Based on these tests, a set of "reference" ring positions was selected. These "reference" positions were then used during subsequent testing over the full range of expected fluid conditions, i.e., steam, loop seal, transition, and water.

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Section 3 .o JUSTIFICATION OF RELIEF VALVE TEST CONDITIONS

3.1 GENERAL

The following sections present the specific conditions under which each power operated relief valve design was tested and justification that the inlet fluid conditions tested adequately represent those for FSAR, Extended High-Pressure Injection, and Cold Overpressurization events in most participating PWR units. Other conditions tested, including the effects of discharge piping on valve operability, are also discussed.

It.should be noted that although Cold Overpressurization events can result in steam, water, or transition flow, in general, tests selected to represent such events included only liquid discharge. This was due to the fact that liquid discharge would result in a PWR if the system were solid which represents a worst case Cold Overpressurization event. In addition, the steam and transition (steamto-water} conditions that might occur during such events are enveloped by those resulting from FSAR and Extended High-Pressure Injection events.

As discussed in section 2.2, valve inlet pressure is considered to have a potential for affecting relief valve performance only when the valve is in the opening or closing process. Three of the relief valve designs tested, (Dresser, Crosby, and Target Rock} open relatively fast, (generally in less than 0.5 seconds}. As a result, even during PWR transients resulting in high pressurization rates, the pressures developed at the inlets of Dresser, Crosby, and Target Rock relief valves while the valves are in the opening process, are only nominally higher than the valve's opening set point pressure.

The highest relief valve opening set point pressure in PWR's utilizing the valves mentioned above is 2450 psig (B&W plants, see reference 1, page 6-1). These valves were tested at inlet pressures slightly above 2450 psig at Wyle Laboratories and within a few percent of this value during Marshall Testing .

3-1

The other relief valve designs tested were asslll1ed to have somewhat longer opening times in the range of two (2) seconds. 1herefore, for very high system pressurization rates the potential ·may exist that inlet pressures an the order of those predicted for uworst case" FSAR averpressurization events may occur while such valves are in the opening process. Test values far maximum inlet pressure were selected for such valves to represent maximum expected pressures far FSAR events in the PWR unit(s) utilizing that specific valve design.

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3.2 DRESSER

Tables 3-1, 3-2, and 3-3 present the conditions under which the Dresser relief valve design was tested during the Marshall, Wyle Phase II, and Wyle Phase III test programs, respectively. Table 3-4 lists the PWR units currently utilizing this valve design. The following sections provide specific justification that for each event type, (FSAR, Extended High~Pressure Injection and Cold Overpressurization) the inlet fluid conditions tested are representative of those expected in these units with the exceptions noted in Table 3-4. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on the Dresser relief valve are discussed.

As shown in Table 3-4, Dresser relief valves are currently utilized only in B&W and Combustion Engineering designed plants. The expected range of inlet fluid conditions for B&W and Combustion Engineering units is provided in references 1 and 2, respectively.

. . 3.2.l FSAR EVENTS

INLET FLUID STATE/PRESSURE/TEMPERATURE

FSAR events result in steam discharge for both B&W (reference 1, page 6-1) and Combustion Engineering (reference 2, Tables 5-1 through 5-21) units. In addition, some of the B&W and CE units listed in Table 3-4, utilize PORV inlet piping which collects a cold water seal at the valve inlet.

Although some FSAR events result in peak pressures higher than the PORV opening set point, testing at the PORV opening set point is considered adequate since the Dresser valve opens quickly (see section 3.1). The highest Dresser relief valve opening set point is 2450 psig (reference 1, page 6-1). The corresponding highest closure set point would be 2425 psig assuming a minimum blowdown of 25 psi.

A total of five (5)* steam tests were performed on the Dresser relief valve with opening pressure above 2450 psig (see Tables 3-2 and 3.3). An additional eleven (11) formal steam tes~s were performed during Marshall testing with opening pressures within 50 psi of this value (see Table 3-1). The maximum closure pressure for these tests was 2320 psig (see Table 3-1) which is within 4.1 percent of the maximum expected value (2425 psig).

*Includes opening on steam during one transition test (21-DR-9 W/W).

3-3

Three cold-to-hot water transition tests were also perfonned on the Dresser relief valve at pressures above 2450 psig to simulate the thennal shock effects of water seal discharge {see Table 3-3). Water seal temperatures as low as 103°F were tested. During one of these tests (24-DR-6W) the signal to close the PORV was given at a pressure of 2345 psig which is within 3.3 percent of the highest expected closure pressure.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Dresser relief valve was tested are considered adequate to represent those expected for FSAR events in the PWR units listed in Table 3-4.

BENDING MOMENT INDUCED

The maximum quasi-steady bending moment induced while the Dresser valve was in the opening or closing process during a Qpreload• test was 25,500 in-lb (see Table 3°3, test 15-0R~SW)a

BACK PRESSURE DEVELOPED

As discussed in section 2.2.4, it is considered important to demonstrate the Dresser relief valve's ability to operate at back pressures at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table 3a4) were perfonned on the Dresser relief valve with valve outlet pressures as high as 760 psia, (see Table 3e3, Test 10-DR .. lS).

3o2o2 EXTENDED HIGH-PRESSURE INJECTION EVENTS


Extended High Pressure Injection events result in full pressure steam to water transition and water (400°F to 650°F) discharge in B&W units (reference l, page 6°3). These events have the potntial for challenging relief valves in' only one CE unit; Maine Yankee (reference 2, Section 6-2). Extended HPI conditions for Maine Yankee include steam to water transition and water (467 to 650°F) discharge (reference 2, Section s.2.1).

A total of six {6} full pressure (greater than 2450 psig) liquid tests were perc

fonned on the Dresser relief valve to represent Extended High Pressure Injection conditions {see Table 3-2, tests DR-5-W, DR-6eW, and DR-7W and Table 3-3 tests ll-DR-4W, 13-DR-7W, and 15-DR-SW). Liquid temperatures ranging from 447°F

3-4

to 647°F were included in thiis test series. ·This range of test temperatures is considered adequate to represent that expected in B&W units and Maine Yankee especial ly since the likelihood of full pressure liquid discharges at temperatures below 450°F is very low (reference 1, Section 4.4.2).

In addition to the water tests described above, a full pressure steam to slightly (-l 5°F) subcool ed water test was perfonned on the Dresser relief valve to represent transition conditions resulting from an Extended High Pressure Injection event, (see Table 3-3,. test 21-DR-SS/W).

Valve closure signals were sent to the Dresser v_alve during these seven (7) tests at inlet pressures ranging from 2105.to 2345 psig depending on the duration of discharge. The highest of these is within 3.3 percent of the maximum closure pressure settings in PWRs (2425 psig).

The inlet fluid conditions under which these seven (7) tests were perfonned are, therefore, considered adequate to represent those resulting from Extended High Pressure Injection events in the PWR units listed in Table.3-4.


See section 3.2.l


·All liquid and transition testing of the ·Dresser relief valve was performed with the same discharge pipe orifice which was demonstrated to develop back pressures between 450 and 500 psia under full pressure steam conditions (see Table 3-3, tests 20-DR-lS and 23-DR-lS). Therefore, the back pressures developed during the seven (7) tests discussed earlier to represent Extended High Pressure Injection events, correspond to those expected under similar flow conditions in in-plant discharge piping systems which develop 450-500 psia under full pressure steam flow conditions.

3.2.3 COLD OVERPRESSURIZATION EVENTS


Cold Overpressurization events in Combustion Engineering units result in maximum inlet pressures of 855 psig (reference 2, Table 5-25) with water and transition steam to subcooled water conditions possible at the relief valve inlet. As noted

3-5

in reference 2, Table 5-25, water temperatures ranging from 100°F to 417°F1are possible with the maximum pressures predicted corresponding to the higher temperature value. Had a lower temperature been asstJ11ed in the NSSS vendor analysis a somewhat lower peak pressure would have been predicted (reference 2, Tables 5-23 through 5-27).

Cold Overpressurization Events in B&W units result in maximum pressures of 550 psig. Transition and subcooled water (338 to 449°F) conditions may exist (reference 1, page 6-4).

The expected transition conditions are represented by the full pressure steam to water transition test performed on the Dresser valve (see Table 3-3, test 21-DRSS/W).

The 855 psig, 417°F water condition is enveloped by tests 0Rm7W and 13-0Rc7W which were full pressure liquid tests at -450°F.

. . The lower end of the expected pressure/temperature range is represented by

test 14-0R-2W~ a 689 psia, 112°F water testG

In all cases discussed above, the PORV closure pressures were well above the core responding closure set points utilized in B&W and Combustion Engineering units {iH=plant closure set points are generally 25 psi below the opening set point) a

The four (4) tests discussed above~ therefore~ are considered adequate to represent the inlet fluid conditions expected during Cold Overpressuri%at1on-events in the PWR units listed in Table 3-4 with the exception noted in the Table.


See section 3a2ol.


Each of the four (4) tests discussed above was performed using a back pressure OFifice which develops 450-500 psia under full pressure steam flow conditionsG Therefore, the back pressures developed during these liquid tests are representative of those expected under similar flow conditions in plants with expected full pressure steam back pressures in this range.

3-6

•

w I

........

• EPRI/ MARSHALL PORV TEST DATA (3)

TABLE 3 • l •As TESTED• MARSHALL TEST MATRIX FOR THE DRESSER JIELJEF VALVE

CONDITIONS AT VALVE OPENING VALVE INLET

TEST TEST . FLUID PRESS. TEMP. NO. TYPE (PSJA) (Of)

STE AH STEAM 2435 (SAT.)

2 STE AH STEAH 2455 (SAT.)

l STEAH STE AH 2440 cm.> 4 STEAM STE AH 2435 (SAT.)

5 STE AH STEAM 2445 (SAT.)

6 STEAll STE AH 2450 (SAT.)

1 STEAH STEAH 2455 (SAT.)

8 STEAM STEAM 2420 (SAT.)

!I STEAM STE AH 2415 (SAT.)

JO STE AH STE AH 2435 (SAT.)


NOTES: (l) Maximum Quasi steady discharge pipe pressure. (2) Not recorded.

JN ACCUMULATOR oollUoN FLUID PRESS. TEHP. (SEC)

(PSIA) (Of)

SAHE AS VALVE INLET 61

26

22

21

24

66

23

27

24

27

(2)

TRANSIENT CONDITIONS ~ALV~ INLET RES • WHEN

HAXIHUM 11) DISCHARG

SIGNAL GIVEN PIPE PRESS. TO CLOSE VALVE (PSIA)

PSIA

2295 415

2235 415

2340 415

2335 415

2335 . 415

2305 175

2335 175

2300 170

2310 170

2330 175

2295 415

(3) Test results are for evaluation tests only. Total of 21 supplementary valve actuation cycles were performed under similar conditions.

•

w D 00

IPRi/ llYLE POR!f llSJ DAiii TMLlr 3l m ~

·.u nsno· llYL[ llllAS[ Bl USf MillSll !!'OR lit[ DRESSER llU!Elf llAn.111£

!HBiiAL (ONOll!ONS ilANSHE~I CONDITIONS IN ACCIHllAiOR iUJ lfALllf UflU 0)

niursur J(SJ l(SJ flUID PRESS. iEKP. lllU!D PA£SS. ,, ... 110. fYPE . (!'SIA) (Of~ (PSIA) (f)

DR-1-S SUM SUM H90 6741 SAHE AS 11ALVf !Ill.EV

DA-HI llAl(A lllAUA WI JU SAHE AS VAL VE 1111.Ei

DR-5-11 MATH MAiflt 2500 646 SAHE AS VALVE INlfi OA-i-V llATEll llAUR Z500 506 SA1E AS VALVf INlU

DH-1-11 lllATER MAUii 2510 4141 W9£ A$ WSllVf INlElr

BIOJ(S:

II) flul~ tonditloni aa ~he ~0Dve 8n!a& 9~~1atei1 prior Qo 1ctuatlon of te§~ ~ai~e. 1) lll&d- Ques I llie&dJf di11cilair911 l!lipe pl1'G!$i\ure. l) Net ~ecorded •

•

i

AS

26

17

16 •

VALVE INLO ""l!IUt J2) PAC SS. Wll(fl DISCHAA SIGNAL GIYUO ?IPf PRESS. TO CLOS( lfALVE (PSIA)

(PSIAI

:ms eo 510 BIO

HOO 290

2120 340

Z»Zlll !JO

IJJASI STEADY tlAXIHIM

PILOT UNE l.P.

(PSIA)

1040

IU

AO

JBO

JJJ

(EP PORV TEST DATA) • ABLE 3-3 . "AS TESTED" WYLE PHASE Ill TEST MATRIX FOR

DRESSER RELIEF VALVE

INITIAL CONDITIONS TRANSIENT CONDITIONS

Valve Inlet

Dack- At Valve Inlet In Accumulator Pressure Maximum Maximum Maximum pressure at Discharge Bending Valve Or if tee Test Closure Pipe Moment Acceleration Area Temp Press. Temp Press. Duration S1gnal Press. lnduced(l)Jnduced

Test No. Test Type (in2) Fluid (Of) (psta) Fluid (Of) (psi a) (seconds) (psla) (psia) (in-lb) (g's)

10-DR-lS ·Steam 4.155 Steam 668 2.503 Steam 669 2.503 15 2,035 760 N/A 6.8.

11-DR-4W Water 6.166 Water 647 2,514 Nater 658 2,514 15 2,338 625 N/A 5.0

12-DR-3W Water 6.166 Water 450 5g9 Water 456 699 15 685 260 N/A 8.2

ll-DR-7W Water 6.166 Water 451 2,4g2 Water 460 2,492 10 2,230 420 N/A 7.4

w I 14-DR-2W Water 6.166 Water 112 689 Water 116 689 10 652 5 N/A 8.6 \0

15-DR-SW Water 6.166 Water 643 2.504 Water 658 2,504 10 2,360 640 25,500 5.3 (Pre load)

16-DR-6W Water 6.166 Water 103 2,500 Water 652 2,500 54 2,320 295 N/A 8.2 Seal Simulation

20-DR-lS Steam 6.166 Steam 657 2,505 Steam 659 · 2.so5 10 2,110 495 N/A 9.2

21-DR-8S/W Transition 6.166 Steam 656 2,496 Water 641 2,496 10 2,360 660 N/A 9.6

22-DR-9W/W Water 6.166 Seal

Water 321 2,490 Water 647 2,490 11 2,310 678 N/A 6.9

Simulation

23-DR-lS Steam 6.166 Steam 657 2,505 Steam 659 2,505 10 2,110 440 N/A 8.9

24-DR-6W/W Water 6.166 Seal

Water 105 2,505 Water 64g 2,505 88 2,360 690 N/A 8.6

Simulation·

Notes: (1) Value shown corresponds to maximum moment applied while valve was in the opening/closing process. (2) Corresponds to t1me PORV was isolated. Actual closure occured at an undetermined t1me after isolation.

TABLE 3-4

PWR UNITS WITH DRESSER RELIEF VALVES

Unit


Oeonee 1, 2il 3

Crystal River 3

Three Mfle Island 19 2

Rancho Seco

Bellefonte 1, 2

Washinqton Nuclear Project 1» 4

Calvert C11ffs 1, 2

Palisades

St. Lucie 1

Maine Yankeei"

M111 stone 2

Fto ~1houn

NSSS Vendor

Babcock and Wilcox

· Combustion Engineering I I I I I I I I I ~

~Justification for conditions resulting from Cold Overpressurization events is not specifically addressed hereino Such justification will be provided as part of this utility•s plant specif1e evaluationo

3 .. 10

•

•

•

•

•

3.3 CROSBY

Tables 3-5, 3-6, and 3-7 present the conditions under which the Crosby relief valve design was tested during the Marshall, Wyle Phase II, and Wyle Phase III test programs, respectively. Table 3-8 lists the PWR unit currently utilizing this valve design. The following sections provide specific justification that for each event type, (FSAR, Extended High Pressure Injection and Cold Overpressurization), the inlet fluid conditions tested are representative of those expected in this unit. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on the Crosby relief valve are discussed.

As shown in Table 3-8, a Crosby relief valve is currently utilized only in a B&W designed unit (Davis Besse). The expected range of inlet fluid conditions for this a&w unit is provided in reference 1.

3.3.1 FSAR EVENTS

·INLET FLUID STATE/PRESSURE/TEMPERATURE ..

FSAR events result in steam discharge in the Davis Besse unit (reference 1, pg. 6-1). In addition, this B&W unit utilizes PORV inlet piping which collects a cold water seal at the valve inlet •

Although some FSAR events result in peak pressures higher than the PORV opening set point, testing at the PORV set point is considered adequate since the Crosby valve opens quickly (see section 3.1). The highest Crosby relief valve opening set point is 2450 psig (reference 1, pg. 6-1). The corresponding highest closure set. point would be 2425 psig ass1JT1ing a minimum blowclown of 25 psi. A total of fifteen* (15) steam tests were perfonned on the Crosby relief valve with opening 'pressures at or above 2450 psig (see Tables 3-5, 3-6, and 3-7). The maximum closure pressure for all steam tests perfonned was 2335 psig {see Table 3-5) which is within 3.7 percent of the maximum expected value.

Three cold-to-hot water transition tests were also perfonned on the Crosby relief valve at pressures above 2450 psig to simulate the thermal shock effects of water seal discharge (see Table 3-7). Water seal temperatures as low as 118°F were tested. During one of these tests (33-CR-7 W/W) the signal to close the PORV was

• *Includes opening on steam during one transition test (31-CR-4S/W).

3-11

given at a pressure of 2285 psig which is within 5.8 percent of the highest expected closure pressure.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Crosby relief valve was tested are considered adequate to represent those expected for FSAR events in the Davis Besse plant.


The maximum quasi-steady bending moment induced while the Crosby valve was in the opening or closing process during a 11preload11 test, was 31,600 in-lb (see Table 3-7, test 26-CR-6S).


As discussed in Section 2.2.4, it is considered important to demonstrate the Crosby relief valve's ability to operate under developed back pressures at least as high as those expected 1n service. Full pressure steam tests (representative of expected FSAR events for the Davis Besse plant) were performed on the Crosby relief valve with valve outlet pressures as high as 545 psig, (see Table 3-6, test CR .. 2-S).

3.3.2 EXTENDED HIGH PRESSURE INJECTION EVENTS


Extended High Pressure Injection events result in full pressure ste~to-water transition and water (400°F to 650~F) discharge fn the Davis Besse plant (reference 1, pge 6-3)a

A total of three (3) full pressure {greater than 2450 psig) liquid tests were performed on the Crosby relief valve to represent Extended High Pressure Injection conditions (see Table 3e6, tests CRc5~W, CRm6-W, and CR-7-W). Liquid temperatures ranging from 446 to 634°F were included in this test series. This range of test temperatures is considered adequate to represent that expected in the Davis Besse plant, especially since the likelihood full pressure liquid discharges at temperatures below 450°F is very low (reference 1, Section 4e4.2).

In addition to the water tests described abovew a full pressure steam to slightly {-15°F) subcooled water test was performed on the Crosby relief valve to represent a transition condition resulting-from an Extended High Pressure Injection event~ {see Table 3-7, test 31-CR-4S/W).

3-12

••

•

•

•

Valve closure signals were sent to the Crosby valve during these four (4) tests at inlet pressures ranging from 1985 to 2298 psig, depending on the duration of discharge. The highest of these is within 5.3 percent of the closure pressure setting assumed for Davis. Besse (2425 psig).

The inlet fluid conditions under which tests were perfonned ~re, therefore, considered adequate to represent those resulting from Extended. High Pressure Injection events in the Davis Besse pl ant.


See Section 3.3.1.


All full pressure 1 iquid testing (with the exception of the water seal simulation tests) of the Crosby relief valve was perfonned with the same discharge pipe orifice which was demonstrated to develop a back pressure of 60 psia under full pressure steam conditions (see Table 3-6, test CR-1-S). Therefore, the back pressures developed during the three (3) liquid tests discussed earlier to represent Extended High Pressure Injection events, correspond to those expected in an in-plant discharge piping system which develops 60 psia under full pressure steam flow conditions. The three (3) water tests performed to simulate water seals (tests 32-CR-5W/W, 33-CR-7W/W and 34-CR-8W/W) were basically 650° liquid tests and can be included in the data base used to represent conditions resulting from Extended High Pressure Injection events. For these liquid tests, the back pressures developed correspond to those expected under 650° liquid conditions in an in-plant discharge pipe which develops a 380 psia back pressure under full pressure steam conditions. This is because the back pressure orifice used during the water seal simulation tests was demonstrated to develop this pressure under steam flow conditions (see Table 3-7, test 30-CR-lS).

3.3.3 COLD OVERPRtSSURIZATION EVENTS

The Davis Besse unit is the only PWR utilizing the Crosby relief valve. This unit does not utilize the PORV as part of its Cold Overpressurization Protection System. Therefore, specific justification of the applicability of the conditions under which the tests were performed to conditions resulting from such events is not requfred •

3-13

w I

II-" .,,.,

lPRI/ IWISHALL PORV l(Sl PATA (l)

IQL!!: :5 - 5

".!Is usn:o~ IWISIW.ll. 1(51 ll!A1Rll fOR lHC d:ROSISll 1m.1£f'

CONOlliON$ l! UALYE OPENING VALVE IHU1 ~· M:CIHll.AfOR oolHIM ·

1ESl iUi RUH! PRUS. n ... nulo PRESS. tcilP. (SEC) HO. nn ~PSIA) C0f) (PSIA) (Of)

a SIEM S1EAll 2495 (SAT.) SAl9E A$ VALVE INLET 06

2 51EAll STEM 24115

J SUM STEM 2455

4 SUM STEM 2455

Ii SUM SlPM 2455

ts SUM STEM 2485

1 SUMI SUM 2485

I STl:M SiEAH 2450

9 SlWi $UM 209i

II $Tl:M sn~ ~465

n lSTIEM SUM HH

NOUS: (21) Ollai1111111 Queal 1te~d1 dlsth1rge ~Qpe pressure.

( ) Hoa ll'li!ctorrdiod.

(SAl.) H

(SRi .J !R

(SAT.) II

ISAi.) II

«SAT.) H

(SAT.B 22

(SA!.B II

!SAll.) 25

(SAW.) H

(SAV.) (2)

VALVE

VRAHSIEH1 COIDl110H5

iftUl.11ill mu:rr.J0 SIGNAL GIVH PIM rans. 10 CLOSE VALVE (PSIA)

PSIA

Zll5 J15

. 2150° JIS

lllli us ZJJO J15

2ll5 J85

2ll5 us U45 us UUi us 2110 es (4)

2150 us !JZS J15

(l) Yest reiulta 11re for 111111111Uon tests Glll~.!f. Vot1l of Jl 1upp1 .. ntar1 valve 1ctu1tlon c1dH 111tra 111treorMd under sl•llar ccondttlons. f4) OownilrrH• prHsure appears to lie 1n111111ous.

• •

w I ......

U1

EPRI/ WYLE PORV TEST DATA TABLE 3 - 6

: "AS TESTED• llYLE PHASE II TEST ~TRIX FOR THE CROSBY llELIEF VALVE

INITIAL CONDITIONS TRANSIENT COHDITIOUS VAL VE INLET ( 1) IN ACCUMULATOR TEST VALVE INLET HAXIHUM (2)

°VRATJOH PRESS. WHEN DISCHARGE TEST TEST FLUID PRESS. TEHP. FLUID PRESS. TEHP. SEC SIGNAL GIVEN PIPE PRESS.

NO. TYPE (PSIA) (Of) (PSIA) ··: . (Of) .... TO CLOSE VALVE (PSIA) IPSIA}

CR-1-5 STEN! STEAH 2150 672 SAME AS VALVE INLET 15 1920 60 CR-2-5. STEAM STEAH 2495 671 SAHE AS VALVE INLET 1 2140 560 CR-3-11 YATER YATER 680 376 SAHE AS VALVE INLET 15 618 125 CR-5-V YATER YATER 2510 634 SAHE AS VALVE INLET 15 2280 155 CR-6-11 WATER WATER 2502 505 SAME AS VALVE INLET 19 2100 315 CR-7-11 YATER WATER 2510 446 SAHE AS VALVE INLET 18 2000 230

NOTES: " (l) fluid condtttons 1t the v1lve tnlet tmnedlately prtor to actuation of test valve.

lZ) Hoxt11111111 Quast steady discharge pipe pressure. l) The 1000 PSIA pressure sensor w1s over-ranged tn thts test.

QUASI STEADY HAlltUt

PILOT LINE 8.P.

(PSIA)

945

)JOOO (3)

200

1i5

OB

661

(ErRft/WYLE PORV TEST DATA~ JAB!.£ l-7

'"AS 1ISVH"' NYLE PHASE Iii TESV MTGWt fOR VHE CROSBY RELIEF VALVE

HNHTRAL COHOIJHONS TRANSIEMT CONDITIONS

Valve inlet

Bacik- At V!!1ve Hllb11et In Accumulator Pressure MHllllllAI Maxi- Maxlmu11 presSUll'e al Discharge Bending Valve Orifh:lll iei;t Closure Pipe Moment Acceleration Are~ l:;t; Preu. Temp Priess. mull'~t@on Signal Press. lnduced(2)1nduced

Test No. Yest lype ~ fluld (ps8a) fluid (Of) (psl.i) (seconds) (psla) (psi a) (In-lb) !~

25-CR-IS Slea111 10.0 Ste11111 656 2.505 Stea111 659 2.so5 10 2,050 (l) NIA 8.0

26-CR-6S Stem !U2 Steam 657 2.505 Ste11111 (Prefo&d)

659 2.505 !O 2,037 (l) n.lioo 6.7

21-CR-2W Malter 9.62 Mmlter IO« 694 Nater 108 6941 10 620 3 NIA 8.2

28-CR-311 W&teir 9.62 Na It er 437 695 Y&lell" 448 695 IO 655 160 H/A 11.6 w e ....

m 29-CR-iS SteMi U.71 Steam 656 2.so5 SlHll 659 2.!505 !O 2,050 14 NIA 6.4

30-CR-IS Steam 6.112 Stea• 656 2,505 Stea• 658 2.505 10 2,060 380 H/A 10.4

l!-CR-45/11 iirans8Uon 8.82 Steam 656 2.sio Nater 649 2.s10 15 2,313 (I) NIA 9.2

l2-CR-5W/lll Ill.Iler 8.82 Nater 469 2,505 Na tell" 646 2.505 15 2,290 560 H/A 8.2 Seal Simta~&Uoo

ll-CR-7M/M Yater au2 Nater 294 2.505 Mater Seal

648 z.505 15 2,JOO 580 ti/A 8.0

SimuhUoo

34-CR-IM/M Waler Ii.ill Mater ne 2.soo Mater 645 2.soo 15 2,290 575 N/A 9.4 Seal S imu\lat Dean

HotH:

i•1 PS-4 was lnoperatl~e dull"lng th9$ test. 2 Value shown corresponds to·11axlmum moment applied! 111he11 valve 111as tn the openlng/c1osigi!ll proceu

• • .

• Unit

Davis Besse 1

•

•

TABLE 3-8

PWR UNITS WITH CROSBY RELIEF VALVE

NSSS Vendor

Babcock & Wilcox

3-17

Comments

177 Fuel Assembly Unit

3.4 TARGET ROCK

Tables 3-9 and 3-10 present the conditions under which the Target Rock relief

•valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-11 lists the PWR units currently utilizing this valve design. The following sections provide specific justification that for each event type (FSAR, Extended High Pressure Injection, and Cold Overpressurization) the inlet fluid conditions tested, are representative of those expected in these units. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on the Target Rock relief valve are discussed.

As shown in Table 3-11, Target Rock relief valves are currently utilized only in B&W designed plants (Midland 1, 2 only). lhe expected range of inlet fluid conditions for B&W units is provided in reference 1.

3.4.1 FSAR EVENTS


FSAR events result in steam discharge for the Midland 1 and 2 units (reference 1, pg. o-1) and these units do not employ a water seal design •

• Although some FSAR events result in peak.pressures higher than the PORV opening set point, testing at the PORV set point is considered adequate since the Target Rock valve opens quickly (see section 3.1). lhe highest Target Rock relief valve opening set point is 2450 psig (reference 1, pg. 6-1). lhe corresponding highest closure set point would be 2425 psig.ass1J11ing a blowdown of 25 psi. A total of seven (7)* steam tests were per~ormed on the Target Rock relief valve with opening

' . ' pressures ~~ove· 2450 psig {see Tables 3-9 and 3-10). lhe maximum closure pressure

.for an· steam tests performed was 2320 psig {see Table 3-9) which is within 4.4 percent of the maximum expected value.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Target Rock relief valve was tested, are consider~d adequate to represent those expected for FSAR events in the PWR units listed in Table 3-11.

*Includes opening on steam during one transition test (l 9-TR-95/W). ·

3-18


The maximum quasi-steady bending moment induced while the Target Rock valve was in the opening or closing process during a "preload" test was 32,900 in-lb (see Table 3-10, test 18-TR-SS).


As discussed in section 2.2.4. 1t 1s considered important to demonstrate the Target Rock relief valve's ability to operate with back-pressures at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table 3-11) were perfonned on the Dresser relief valve with valve outlet pressures as high as 460 psig (see Table 3-9, test 5).


INLET FLUID STATE/PRESSUREGEMPERATURE

Extended High Pressure Injection events result in full pressure steam to water transition and water (400 to 640°F) discharge in the Midland l and 2 units (reference 1, pg. 6-3).

A total of four (4) full pressure (greater than 2450 psig) liquid tests were performed on the Target Rock relief valve to represent Extended High Pressure Injection conditions (see Table 3°10, tests 4-TR-SW, 6-TR·4W» a~TRmSW, and 9-TR-6W)G Liquid temperatures ranging from 451°F to 648°F were included in this test ser1eso 'Otis range of test temperatures 1s considered adequate to .represent that expected in the Mi.dland 1 and 2 units, especially since the likelihood of

full pressure liquid discharges at temperatures below· 450°F fs very low (reference 1, section 4.4.2)o

In addition to the water tests described above, a fu11 pressure steam to slightly (-15°F) subcooled water test was performed on the Target Rock relief valve to represent a transition condition resulting from an Extended High Pressure Injection event (see Table 3-10, test 19°TRm9S/W).

Valve closure signals were sent to the Target Rock valve during these five (5) tests at inlet pressures ranging from 2181 to 2305 psig depending on the duration of discharge. The highest of these is within 5.0 percent of the maximum closure pressure setting assuned for the Midland units (2425 psig).

3-19

•

•

•

•

•

The inlet fluid conditions under which these five (5) tests were perfonned are, therefore, considered adequate to represent. those resulting from Extended High Pressure Injection events in the PWR units listed in Table 3-11.


See Section 3.4.1.


All liquid and transition testing of the Target Rock relief valve was perfonned with the same discharge pipe orifice which was denonstrated to develop back pressures between 315 and 330 psia under full pressure steam conditions (see Table 3-10, tests 1-TR-lS, 2-TR-lS, 17-TR-lS, and 18-TR-SS). Therefore, the back pressures developed during the five (5) tests discussed earlier to represent Extended High Pressure Injection events, correspond to those expected under similar flow conditions in an in-plant discharge piping system which develops 315-330 psia under full pressure steam flow conditions.



Cold Overpressurization events in the Midland 1 and 2 units result in steam discharge at a maximum pressure of 550 psig and steam followed by water discharge at a maximum pressure of 360 psig. The range of water temperatures possible for such events is from 415°F to 426°F (reference 1, Table 6-3).

The expected transition condition is enveloped by the full pressure steam to water transition test perf~nned.(see Table 3-·10, test 19-TR-9S/W) •

. . '

The expected liquid pressures/temperatures are enveloped by tests 3-TR-3W (715 psi a, 447°F water test) and 5-TR-2W (690 psi a, 114°F water test).

During all tests discussed above, the PORV closure pressures were well above the opening set point utilized in Midland 1 and 2 {360 psig).

The three (3) tests discussed above, therefore, are considered adequate to represent the inlet fluid conditions expected during Cold Overpressurization events in the PWR units listed in Table 3-11 •

3-20

BE~OING MOMENT INDUCED

See Section 3.4.1.


Each af the three (3) tests discussed above was per-fanned using a back pressure orifice which develops 315-330 psia under full pressure steam f1 ow ~onditions .. Therefore~ the back pressures developed during these liquid test~ are representa~ tive of those expected under similar flow conditions in plants with expected full pressure steam back pressures in this rangea

3-21

•

•

•

•

w I

N N

--------

EPRJ/ MARSHALL PORV UST DATA (l)

TABLE l • 9 "AS TESTED• MARSHALL TEST MATRIX FOR THE TARGET ROCK RELIEF VALVE

CONDITIONS AT VAL¥£ OPENING

VALVE· INLET TEST TEST FLUID ~:~~~;: TEMP.

NO. TYPE (Of)

1 STEM STEAM 2475 - (SAT.)

2 STEAM STEM 2455 (SAT.)

J STEM STEAM 243S (SAT.)

4 STEM STEM 244S '(SAT.)

5 STEM STEAM 2455 '. (SAT.)


1 STEAi-i STEM 2445 (SAT.)

8 STEAM . STEAM 2425 (SAT.) g STEAM STEAM 2455 '(SAT.)

10 STEAM STEM ·:z.75 (SAT.)

II STEM STEM 2455 ,<SAT.)

NOTES: (I) tlllalmum Quast steady discharge ptpe pressure. (2) Not recorded.

IN ACCUHULATOR oulliloN FLUID PRESS. TEMP. (SEC)

(PSIA) (Of)

SAME AS VALVE INLET 68

31

59

25

25

60

28

26

zg

30

(Z)

TRANSIENT CONDITIONS

~ALY~ INLP RES • 1111 N MAXltllH p> DIS HARG SIGllAL GIVEN PIPE PRESS. TO CLOSE VALVE (PSIA)

PSIA

2310 465

2310 465

2295 465

U20 465

2ll5 475

2295 155

2305 155

2295 155

2335 155

2335 155

2325 455

13) Test results ire for evaluation tests only. Tot1I of 23 supplementary v1lve actuation cycles were performed under slmllnr conditions.

(EPRl/WYLE PORV TESJ DATA) 1MB8..E JJ-10

•As VIESYIEO"' WYLIE fl'!IASIE m TESJ MATRllt fdlfi il!E iAllGIEV ROCK RELIEF VALVE

RHITiAL COHOHTHONS TRANSIENT to~_HJ.!l~L. _____ Va Ille In Bet

Back- At lfa1ve &ntea IDll Accumulator Pressure Haxl111111 Maxlllllll H111tm1111 preuGBre at Discharge Bending Valve Orifice Jed Closure Pipe Moment Acceleratlon Area iemp l?reu. Temp Press. Dura~lon Segn11 Press. lnduced(l)lnduced

_Test Ho. les!_.!ll!.. ( tn2.L_ !_Md ~ Je~!l fJutaB j~LJ- .Ce!!~} h~.!:on~ h~~~~L .<.e~t.!L_ H~L (g•st_ ____

U-T.R-IS Steam 9.62 Ste1:1111 660 252R Stea111 610 2521 1 2132 320 N/A 5.0

2-iR-IS Steam !UZ Steam 669 2504 Steam 670 . 250'11 1 2134 330 NIA 6.6

3-TR-JW Water 9.62 Water 447 715 water 454 715 70 639 170 H/A 3.3

4-TR-511 Hater !U2 Miiter 645 2516 Willer 653 2536 iS 2293 450 N/A 3.6 w D

N w 5-lR-211 Haler 9.62 Ma~er U4 690 Nater 114 690 iO 616 N/A 11.2

6-!R-4W Maier 9.62 Naae1r 451 2500 Nater 461 2508 10 2196 l95 H/A 7.4

1-1R-1M Water !il.ii2 Water Ill 2505 Water 656 2505 21 2271 520 N/A 5.6 SHB SimulatOon

O-VR-5W Waler 9.62 Na It.er 648 2494 Waler 658 2494' 10 2120 4JO N/A ~.-t

9-iR-611 Water !11.eiZ Mate~ 645 2490 Y.aier 657 2490 llO· 2302 425 H/A 8.4

11-iR-iS Sleam 9.62 Steam 657 251.0 saeil1111 659 2510 10 2028 325 ff/A 4.9

Ill- iR-85 Stea1111 9.62 S~em.i 656 2505 Slelil1111 650 2505 m 2020 315 32.900 ].JI (Prelooid)

i9-iR-9S/W iransltiol'I !U2 Steam 656 2500 Water 642 2500 iO 2ln0 435 Hiii. 5.2

Hoaes: (I) llalue silow11 corresponds to maiilm11111moment applied whllle val11e was In openlng/cioslng process.

• •

•

•

Unit

Midland 1, 2

Table 3-11

PWR UNITS WITH TARGET ROCK RELIEF VALVES

3-24

NSSS Vendor

Babcock & Wilcox

•

3.5 CONTROL COMPONENTS

Tables 3-12 and 3-13 present the conditions under which the Control Components relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-14 lists the PWR units currently utilizing this valve design. The following sections provide specific justification that the inlet fluid conditions tested are representative of thos.e expected for FSAR and Extended High Pressure Injection events in these units. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on the Control Components relief valve are discussed.

It should be noted that the Control Components relief valve utilizes air pressure to open and to assist the spring in closing the valve (all the other air operated PORVs tested utilize a spring only for closure). For the purpose of this program, it is necessary only to demonstrate the operability of the PORVs when supplied with normal air and electrical inputs. However, to fully detennine the operational characteristics of this valve design, several tests were performed with a "faHed air" supply during closure to assess the capability of spring alone to close the valve. Since the air assist normally supplied on closure increases the valve 1 s ability to close, credit can clearly be taken for results observed during "failed air" on closure tests when developing justification for the range of conditions tested.

As shown in Table 3-14, Control Components relief valves are currently utilized only in Westinghouse designed units (McGuire 1, 2 and Catawba 1, 2). The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.5.1 FSAR EVENTS

INLET FLUID STATE/PRESSURE/:rEMP.ERATURE

FSAR events result in steam, transition, and water discharge in the McGuire and Catawba units (reference 3, Tables 5-1 and 5-2*). These units do not utilize a water seal design.

*Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses perfonned to obtain these conditions assumed that the PORVs were not operable. However, if the PORVs are operable conditions representative of those presented can be expected at the PORV inlets as well .

3-25

As discussed in Section 3.1, an assuned opening time in the range of two (2) seconds for the Control Components relief valve, necessitates denonstration of operability at inlet pressures up to the maximum predicted for FSAR

Overpressurization events.

The McGuire and Catatt'ba units are of the 4-1 oop design. The maximum expected PORV inlet pressure for FSAR (steam discharge) events in 4-loop Westinghouse unitsll is 2532 psia (roeference 3, Table 5-l)o The maximum pl"essure attained during FSAR

· events leading to transition or water discharge in these units is 2507 .7 psi a

(reference 3, Table 5-2). From the same table, the range of expected liquid temperatures is defined as 611.9 to 631.3°F.

A total of eight (8) steam tests were perfonned on the Control Components relief valve with inlet pressures well is excess of the maximum predicted (2532 psh) for

FSAR events (see·Table 3-13, tests 35, 36, 42, 45, 47 9 49, 50, and Sl)a In addi .. tionll a steam (2530 ps1a) to 647°F water transition test (see Table 3-13» test 46) was performed to address possible transition conditions res.u1ting from FSAR events. To represent expected liquid conditions resulting from FSAR events in these units, several full pressure liquid tests were perfonned (see Table 3-13, test 41, 43, and 44). Inlet opening pressures for all these tests were above the maximum expected during FSAR events resulting in transition or liquid discharge (2507&7 psia) and liquid temperatures ranged from 633@F to 645°Fe

PORVs in Westinghouse units have opening set points of 2350 psia (reference 3, Table 3°2)a The blowdown setting for Westinghouse PORVs is nominally 20 psi (r~ference 3, page 2°7)a Therefore, the expected PORV closure pressure in Westinghouse units is 2330 psiac During all tests discussed above, the signal to close the PORV was given at inlet pressures at or above this value except during test 36e The closure signal for this test was given at a pressure of 2280 .psi a which is within 2.2 percent of the expected valueo

Based on the above, the inlet fluid conditions (state~ pressure, and temperature) under which the Control Components relief valve was tested are considered adequate to represent those expected for FSAR events in the PWR units listed in Table 3cl4.

3-26

•

•

•


The maximum quasi-steady bending moment induced while the Control Components valve was in the opening or closing process during a 11 preload11 test was 39,000 in-lb (see Table 3-13, test 47-CC-3S).


As discussed in Section 2.2.4, back pressure is not considered a primary parameter affecting the operation of the air operated PORVs tested. Therefore, although back pressures in the range of those expected in the Catawba and McGuire units were tested, plant-specific justification that the tested values exceed those expected is not required.

3e5.2 EXTENDED HIGH PRESSURE INJECTION EVENTS


Extended High Pressure Injection events result in steam followed by water discharge through relief valves in 4-loop Westinghouse units including the catawba and McGuire units (reference 3, Section 5-3). Liquid discharges at the PORV s.et point (2350 psia) with temperatures in the range of 565°F to 569°F are predicted for this event (reference 3, Tabl~ 5-3) •

To represent the.conditions resulting from such events, one steam-to-water t.ransition and several water tests were performed. The 1 iquid tests (tests 38, 41, 43, and 44) were performed at opening pressures exceeding 2350 psia and temperatures ranging from 440°F to 645°F.

The transition test (test 46) was performed with an opening pressure ~bo.ve . . '

2350 psia with a liquid temperature of 647°F. C~~sure pressures for all tests discussed above exc!!eded. that ~xpected · fo the Catawba and McGuire units

· (2330 psia) except for tests 38-CC-5W and 46-CC-8S/W. Test 38 was a full pressure, 440°F liquid test. The closure signal was given at 2180 psia during this test which is within 6.2 percent of the expected value. Test 46 was a full pressure transition test. The closure signal was given at a pressure of 2320 psia during this test which is within 0.5 percent of the expected value.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Control Components relief valve was tested are considered adequate. to represent those expected for Extended High Pressure Injection events in the units listed in Table 3-14.

3-27

BENDING MOMENT INDUCED See Section 3.5.1.

BACKPRESSURE DEVELOPED See Section 3.5.1.

3.5.3 COLD OVERPRESSURIZAiION EVENTS The range of expected fluid conditions for Cold Overpressurization events in the units listed in Table 3-14 was not provided in the Westinghouse '0Plant Conditions Justification Report" (reference 3) for the units listed in Table 3-14 (see section 1.3 for discussion). Therefore, justification that the conditions under which the Control Components relief valve was tested represent those expected during such events in these units is not presented and w111 be included as part of the plantG specific evaluations.

3-28

•

•

w I

N ID

EPRI/ HARSHA,., PORV TEST DATA (3)

TABLE 3 - 12 •As nsuo• MARSHALL TEST MATRIX FOR THE CONTROL COMPONENTS RELIEF VALVE

CONDITIONS AT VALVE OPENING TRANSIENT CONDITIONS VALVE INLET

TEST TEST FLUID PRESS. TEMP. NO. TYPE (PSIA) (Of)

STE AH STEAM 2455 (SAT.)




5 STE AH STEAM 24J5 . (SAT.)

6 STEAH STEAM 2435 (SAT.)

7 STEAH STEAi! 2405 (SAT.)

8 STEAH STEAM 2395 (SAT.) 9 STEAM STEM 2455 (SAT,)

10 STE AH STEM :ms (SAT.)

11 STEAM ST£AH 2435 (SAT.)

12 STE AH STE AH 2435 (SAT.)


14 STEAM STEAM 2405 . (SAT.)

15 STEAH STEAM 2395 (SAT.)

16 STE AH STEAM Z405 (SAT.)

17 ST£AH STEAM 2415 (SAT.)

NOTES: (1) Hoilmum Quesl steady discharge pipe pressure: (2) Not recorded. ·

.IN ACCUMULATOR oulUJ°" JALVl INLET RES • WHEN FLUID PRESS. UHP. (SEC) SIGNAL GIVEN

(PSIA) (Of) TO CLOSE VALVE PSIA

SAKE AS VALVE INLET 67 2155

30 ', 2175

31 2175

37 2175

37 2195

24 2165

• ZS 2145

:21 2145

155 2155

18 2095

:ro 2175

118 2180

l?3 2170

117 2155

119 2150

118 2155

{2) 2170

umw.:Gp> PIPE PRESS.

(PSIA)

615

615

615

615

615

615

615

615

220

215

215

215

215

215

215

215

615

(l) Test results ire for evaluation iests only. Total of 33 supplementary valve actuation cycles were performed under similar conditions.

(EPRl/WYLE PORV TEST DATA) TABLE 3-13

------- "AS iESn:o• WYLE PHASE Ill TEST HATRIX fOft THE COHiROL COMPONENTS RELIEF VALVE

INmM. COHDITIONS(I) TRANSIENT CONDITIONS

V1'ive Inlet

Valve Inlet Acc1111111111tor Pressure Max llllilll Nut .. Mox 1111111 It Dlsch1rge Bendtne Wo1ve

Test Closure Pipe Modena Acceler11tlon Temp f!l'eH. I~ Press. Du!l'atlon Slgnol Press.(2) lnduced(l)lnduced

!!Sl No. Vest Type nutd ~ Jru!l. fluld (psl11) (seco'!c8s) (psla) (psto) (In-lb) (g's)

35-CC-!S Ste4111 Sten 68] 2.760 Slel!S 676 20 760 Ii 2.374 468 N/A 12.5

l6-CC-2S' Stea11 Ste• 683 2.150 SteU. 678 2,751)) 6 2.200 416 H/A 5.3

(filled Air) l1-CC-3S Steam Stelllll 1510 2.535 Ste• 665 2.535 ~ 2.no 311 rn.ooo 6.2

(Pre1ead0

fllied IUr)

w 38-CC-5W Wiler ldl!lt.e!I' 440 z.536 V1tew- 449 2.sl6 5 z.aoo 400 N/A U.5 D (Failed A9w-) w 0 39-CC-6W Water W11ter IOl 4175 W1le!I' 107 475 10 42! 15 H/A 9.5

(failed Alw-) 40-CC-4W Walen- Ymte;- 392 524 Mltew- 397 524 15 Ul 145 N/A 5.1

(Fil Uedl, AflrrD 41-CC-7" Water Illa terr 633 2.535 liio11ter 654 2,535 25 2.342 480 N/A 6.l

(Failed 'Ud 42-CC-IS Steam Stem. 683 2.760 Ste!!!I 678 2,760 Qi 2,340 450 H/A 11.3

O-CC-7W Water Nater 645 2,538 Water 649 i 2,538 6 2,337 510 N/A 7.7

44-CC-7W Wiler Wate;- 644 Z,540 Nater 656 2,540 48 2.330 490 N/A 5.5

(hl'ied Air)

45-CC-IS Steam Stea 683 2,760 Steam 678 Z,760 4 2,408 45] NIA ~0.7

li6-CC-8S/W ironslUon Stem 664 2.530 Nate1r 641 2,530 ' 2,320 500 H/A ll.O

(filled Aid

41-CC-lS Steam Steam 68) 2.760 Steam 678 2.760 " 2,410 475 l!i'J,000 11.4

(Preh!aci Faileci A9d

•

• • • .

(EPRl/WYLE PORV TEST DATA) TABLE 3 - 13 (cont'd)

•AS TESTED• WYLE PHASE Ill TEST MATRIX FOR THE CONTROL COMPONENTS RELIEF VALVE

(CONT'D)

INIT!~! CO~!I!ON~~ !.) .. ·--- ·--·- --~ ......... _____ . ______ lRANS!!._lff CONDI!!!!~----

Valve Inlet

Valve Inlet Accumulator Pressure Maximum Maximum Maximum al Discharge Bending Valve

Tesl Closure Pipe Moment Acee lerallon Temp Press. Temp Press. Duration Signal Press.(2) Induced Induced

Test Ho. TesL!IJ!! .fluid jOfl .l~!!l. .fluid . jOfl (p~f.!! _t~econds) (p~!!L_ (e~!!l_ (In-lb) (,_'..!L_

48-CC-9W/W Water Waler 135 2,540 Water 648 2,540 6 2,340 515 ff/A 3.9 Seal Simulation

49-CC-2S Steam Steam· 683 2,760 Steam 678 2,760 1 2,380 440 ff/A 11.4 (failed Air)

w 50-CC-3S Steam Stea11 . 683 2,760 Steam 678 2,760 4 2,410 460 38,000 6.4 I w

(Pre load) ......

51-CC-lS Stea1a Steam 683 2,760 Steam 679 2,760 3 2.450 473 36,800 7.3 (Pre load failed Air)

(I) GN2 PORV Actuation Ullage pressure for al! tests was 85 (!. 5) pslg. (2) No Back Pressure orifice was used In the' Control Components PORV testing. (l) Value shown corresponds to maximum moment applied while valve was In opening/closing process.

Table 3-14

PWR UNITS WITH CONTROL COMPONENTS RELIEF VALVES

Unit(l)

V. B. McGuire 1, 2 (OAP, DBP)

Cata~a 1, 2 (DCP 1 DDP)

NSSS Vendor

Westinghouse

Westinghouse

Comments

Four loop units

Four 1 oop unit$

CllJustification for conditions resulting from Cold OVerpressurization events f s not presented for these units. 'This· justification will be provided as part of these unit.s' plant specific evaluations.

I I

3-32

•

•

•

•

•

3.6 MASONEILAN

Tables 3-15 and 3-16 present the conditions under which the Masoneilan relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-17 lists the PWR units using or planning to use the Masoneilan relief valve design. The following sections provide specific justification that the inlet fluid conditions tested are representative of those expected for FSAR, Extended High Pressure Injection, and Cold Overpressurization* events in these units. In addition, discharge piping effects on valve operability (back pressure and bending moment) imposed on the Masoneilan relief valve are discussed.

As shown in Table 3-17, Masoneilan relief valves are currently utilized only in Westinghouse 2, 3, and 4 loop units. The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.6.l FSAR EVENTS


FSAR events can result in steam, transition, and water conditions at the inlets of relief valves in some of the units listed in Table 3-17 (reference 3, Tables 5-1 and 5-2**). In addition, some of these units currently utilize a cold water seal design.

A"s discussed in Section 3.1, an assumed opening time in the range of two (2) seconds for the Masoneilan relief valve, necessitates demonstration of operability at inlet pressures up to the maximum predicted during FSAR Overpressurization events. The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in Westinghouse units, is 2573 psia {reference 3, Table 5-1)_•; The .... · maximum pressure expected during FSAR events resulting in transition or.·liquid discharge, in the unit~ listed in Table 3-17, is 257S·psia {r~fere~ce 3, Table 5-2). From the same table, the range of expected liquid temperatures for these units is defined as 634 to 672°F.

*Cold Overpressurization conditio.ns justified only for plants as noted in Table 3-17. Such justification for the remaining units having this valve design will be provided as part of those utilities' plant specific evaluations.

**Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses performed to obtain these conditions assumed that the PORVs were not operable. However, if the PORVs are operable, conditions representative of those presented can be expected at the PORV inlets as well.

3-33

One (1) water seal simulation and two (2) steam tests were perfonned on the Masoneilan relief valve with inlet opening pressures (2765, 2758, and 2640 psia) well in excess of the maximum predicted (2573 psia) for FSAR events (see Table 3-16, tests 61, 52 and 53).

To address transition and liquid conditions resulting from FSAR events in these units, one test of each type was perfonned. The transition test {see Table 3al6, test 60) was perfonned at a pressure of 2535 psia with 647°F water following the steam initially at the valve inlet. The liquid test (see Table 3-16, test 59) was perfonned with an opening pressure of 2533 psia and a temperature of 648°F. For both tests the liquid temperatures tested (647 and 648°F) were representative of those expected (634 to 672°F). The opening pressures tested {2535 and 2533) were slightly below the maximum predicted value (2575 psia). However, the NSSS vendor analysis asstJned that the PORVs were not operable and that the safety valve opened

. 75 psia above its set pointe The tested pressures are considered representative of the peak pressure that would have been predicted had operation of the relief valves been taken into account in the analysiso

PORVs in Westinghouse units have opening set points of 2350 psia during power operation (reference 3, Table 3-2). The nominal PORV blo~own setting in Westinghouse units is 20 psi (reference 3, page 2-7). Therefore» the closure pressure for- relief valves in Westinghouse units is 2330 psia. During all tests d·iscussed above 11 the signal to close the Masonei1an relief valve was given at inlet pressures exceeding this value.

Based on the above, the inlet fluid conditions.(state, pressure, and temperature) under which the Masoneilan. r:-elief valve was tested, are considered adequate to represent those expected for FSAR events in the PWR units listed in Table 3-170

.. '


The maximum quasimsteady bending moment induced while the Masoneilan valve was in the opening or closing process during a 11prel oadllil te!\it was 35 ,600 i n-1 b (see Table 3~169 test 53..MN-2S).

•

•

•


As discussed in Section 2.2.4, back pressure is not considered a primary parameter affecting the operation of the air operated PORVs tested. lherefore, although back pressures in the range of those expected in the units listed in Table 3-17 were tested, plant-specific justification that the tested values exceed those expected i's not required.



Extended High Pressure Injection events result in steam followed by water discharge through the relief valves in some of the Westinghouse units listed in Table 3-17 (reference 3, Section 5-3). Liquid discharges at the PORV opening set point (2350 psia) with temperatures in the range of 498 to 569°F are predicted (reference 3, Table 5-3).

To represent the conditi ans resulting from such events, one transition and several liquid tests were performed. lhe liquid tests {see Table 3-16, tests 54, 59, and 62) were performed at opening pressures exceeding 2350 psi a and temperatures· ranging from 327 to 648°F. As discussed in Section 3.6.1, a transition test {see Table 3-16, test 60) was also performed with an opening pressure above 2350 psia and a liquid temperature of 647°F. Closure pressures for all tests discussed above exceeded that expected in the units listed in Table 3-17 with the exception of test 54-MN-4W. The closure pressure for this test {2150 psi a) was within i.8 percent of the maximum expected closure pressure (2330 psia).

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Masoneilan relief valve was tested, ?re considered adequate to.· ·

' • • I

represent those expected for Ex~ended High Pressure' Injection events 1n the units 1 isted in Table 3-17.


See Section 3.6.l•

3-35

BACl< PRESSURE DEVELOPED

See Section 3.6.l.


The range of expected fluid conditions for Cold Overpressurization events·was provided via the Westinghouse Plant Conditions Justification Report (reference 3) only for those units as noted in Table 3-17, i.e .. D only for the Sequoyah 1 unit (see Section 1.3 for discussion). Therefore, justification that such conditions were addressed in the testing of the Masoneilan relief valve for the other units listed in Table 3-17 will be included as part of their plant-specific evaluations.

INLET FLUID STATE/PRESSURE/TEK>ERATURE

The full range of potential Cold Overpressurization conditions for all Westinghouse units evaluated are depicted 1n Figure Sml of refe~ence Jo As discussed in Section 5.4 of reference 3, steams transition5 and subcooled wat~r conditions occurring over a large range of pressure and t~era~ure are possible during such events in the Westinghouse units evaluated· ..

Fran Figure 5-1 of reference 3, i~ is noted that full pressure (2350 psia) liquid · discharges are expected over a temperature range of 250°F to 650°F. Based on this figure and discussions 1n Section 5.4 of refeFefice 39 lower pressure (300 psia) liquid discharges are expected over a 1 iquid temperature range of 100 to 400°F o

The fu11 pressure (greater than 2350 psia) steam and transition tests previously discussed (Table 3-16, tests 52, 53, and 60) envelop expected steam and transition· conditions resulting fran such events.

L1 quid tests over a range of pressures and temperatures representative of those shown in Figure Ssl, reference 3, were also perfonnedo Full pressure (greater than 2350 psia) tests with liquid temperatures ranging from 327'i' to 648°F were perfonned {see Table 3-16, tests 54, 59» and 62)o In addition~ reduced pressure (-675 psia) tests with liquid temperatures ranging from 101 to 445°F were perfonned (see Table 3-16, tests 55~ 56~ 57, and 58).

'*Although the 327°F value 'is greater than the lowest possible temperature {at full pressure) in Westinghouse units (250°F), it is representative of the minimum expected temperature in t~e Sequoyah 1 unit.

I i

3-36

•

•

Valve closure pressures for all full pressure liquid tests discussed above exceeded the expected closure pressure (2330 psia) except test 54. The closure pressure for this 444°F water test was 2150 psia which is within 7.8 percent of the maximum expected value. Valve closure pressures for all reduced pressure tests were well above the expected opening pressure (300 psia) corresponding to the lower end of the expected liquid temperature range (100-400°F).

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Masoneilan relief valve was tested are considered adequate to represent those expected for Cold Overpressurization events in the Sequoyah 1

unit.


See Section 3G6.1.

BACK PRESSURE DEVELOPED See Section 3.6.1 •

3-37

----~--

w e w IOO

Ji'RI/ HARSllALL PORV UST llATA (JJ

JlllLlt J - H "M 11HUO" MARSHALll. UST KUlllU iOll 111£ MASOHEUM Rf!..IEF "AUE

'ONDlilONS AV WAlVE OPENING ilANSIEHl CDHOITBONS

WALY( ftNLET • IN ACCUIU.AlOR oullUOH HHl. 111hH UST UH Fl.UID PRESS. n ... FLUID mss. TEtli. (SlCI 516HAL GIVH lllO. iYfE (PSIA) (Of» (l!'SIA) (Of) TO CLOSE VALVE

PSIA

I SU:Alt SWiii HIS (SAT.) SANE AS WAL V£ lllLEi H 2205

2 SUM STEM 2415 (SAT.»

J SUM 5Tf All 1455 (SAi.)

• STEM STEAll 2455 (SAT.)

I SUM STfAll 1H!l5 CSAT.)

6 STl£Nll STEM iH85 (SAT.D

., SUM STltA!!I 2505 (SAT.)

I 511EM SU:AMl !4175 (SAV.i

9 STEM SUM 2455 (SAV.»

BO STEM SUM 1~6$ (SAT.)

u STUN STEM i495 (SAY.)

MOill:S:

i!) 1!1Aai11111111 Quasft 1teady ~lsch1rgE plPQ pressure. (g) !lot !TtCOrdelS.

~z :ms :u 2205

" 2215

H 2245

IH 2195

31 22:J5

J5 1215

4D 2205

36 2215

U) 2225

1Rlll:J0 PIPE l'A£S5.

(PSIA)

525

115

IJS

5J5

545

115

B85

115

!15

185

54$

iJ) Test results nre for 1v11uat~OI! &e1tl on11. iot1I of 19 8upplet111ni1r1 ~•ive actuation cycleg were ijllrfol!"lllld under Gl•ll1r candltlona.

• • (EPRl/WYLE PORV TEST DATA) TABLE 3 - 16

•AS TESTED" WYLE PHASE Ill TEST MATRIX FOR THE MASONEILAN RELIEF VALV[

INITIAL CONDITIONS TRANSIENT CONDITIONS ·----··------- ------- - ... ····-··--------- -----------------·-· ·--···---··-···- ·- ..

Valve Inlet

At Valve Inlet In Accumulator Pressure Maximum Maximum Maximum A1r at Discharge Bending Valve Accum. Test Closure Pipe Moment Acceleration Press. Temp Press. Temp ~ress. Duration Signal Press.(2) 1nduced(4)1nduced

Test No. _!est Type (pslgHl) Fluid J~f.J _(p~-!_~ Fluid J.~U _(ps~ (seconds) !_p~~!l__ Ie~!_a_) _ P!'-.1.~) _ <11~~)_ __ -------

52-HN-lS Steam 54 Steam 683 2,765 Steam 679 2,765 7 2.415 358 N/A 3.3

53-MN-2S Steam ·54 Steam 683 2.758 Steam 678 2,758 8 2.370 346 35.600 3.1 (Pre load)

54-MN-4W Water 54 Water 444 2.530 Water 448 2.530 12 2.150 396 N/A 9.1 w I w \()

55-MN-JW Water 54 Water 445 678 Water 450 678 11 652 153 N/A 1.9

56-MN-5W Water 56 Water 104 675 Water 116 675 11 593 38 N/A 4.5

57-HN-3W Water 58 Water 444 674 Water 445 674 11 640 156 N/A 2.0

58-MN-5W Water 59 Water 101 675 Water 101 675 20 570 14.7 N/A 5.0

59-MN-6W Water 59 Water 648 2.533 Water 650 2.533 9 2.355 435 N/A 3.8

60-MN-7S/W Transition . 59 Steam 670 2,535 Water 647 2,535 9 2,350 425 N/A , 4.0

61-MN-811/W Water 59 -Water 115 2.640 Water 656 2.640 17 (3) 450 N/A 5.9 Seal Simulation

62-MN-9W Water 59 Water 327 2.672 Water 325 2.672 6 2.340 112 N/A 9.6

TP.BLE 3 ~ 16 (Cont'd)

Notes:

(1) All values ! l psic

(2) No back pressure orifice used during Masoneilan valve testing.

(3) Inlet pressure transducer failure.

(4) Value shown CDM'esponds to the maximum moment applied while the valve was _in the opening/e1osing process.

••

•

Table 3-17

PWR UNITS WITH MASONEILAN RELIEF VALVES

.Unit NSSS Vendor Comments

•

Beaver Valley 1 (DLW)l

Donald C. Cook 1, 2 (AEP, AMP)l

Diablo Canyon 1, 2 (PGE, PEG)l

Seauoyah 1 (TVA)

North Anna 1, 2 (VRA, VGB)l

Kewaunee (WPS)l

Westinghouse

Westinghouse

Westinghouse

Westinghouse

Westinghouse

Westinghouse

Three loop unit

Four 1 oop units

Four 1 oop uni ts

Four 1 oop unit

Three loop units

Two loop unit

lJustification for conditions resulting from Cold Overpressurization events is not oresented for these units. Such justification will be provided as part of these utilities' plant-specific evaluations •

3-41

3.7 COPES-VULCAN (17-4PH PLUG AND CAGE)

Tables 3-18 and 3-19 present the conditions under which the Copes-Vulcan (17-4PH Plug and Cage) relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-20 lists the PWR units using or planning to use the Copes-Vulcan (17-4PH Plug and Cage) relief valve design. The following sections provide specific justification that the inlet fluid conditions tested are representative of those expected for FSAR and Extended High Pressure Injection events in these units. In addition, discharge piping effects on valve operability (backpressure and bending moment) imposed on the Copes-Vulcan (17-4PH Plug and Cage) relief valve are discussed.

As shown in Table 3-20, Copes-Vulcan (17-4PH Plug and Cage) relief valves are currently utilized only in Westinghouse 2, 3, and 4-loop units. The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.7.1 FSAR EVENTS


FSAR events can result in steam, transition, and water conditions at the inlets of relief valves in some of the units listed in Table 3-20 (reference 3, Tables 5-1 and 5-2*). In addition, some of these units utilize a cold water seal design.

As discussed in Section 3.1, an assumed opening time in the range of two (2) seconds . .

for the Copes•Vulcan relief valve, necessitates demonstration of operability at in-let pressures up to the maximum predicted during FSAR Overpressurization events. · The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in Westinghouse units, is 2573 psia (reference 3, Table 5-1). The maximum pressure expected during FSAR events resulting in transition or liquid discharge, in the units listed in Table 3-20, is 2575 psia (reference 3, Table 5-2). From the same table, the range of expected liquid temperatures for these units is defined as 654 to 6580 F.

*Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses performed to obtain these conditions assumed that the PORVs were not operable. However, if the PORVs are operable conditions representative of those presented can be expected at the PORV inlets· as well.

3-42

One (1) water seal simulation and two (2) steam tests were perfonned on the Copes Vulcan (17-4PH Plug and Cage} relief valve with inlet opening pressures (2745, 2745, and 2715 psia) well in excess of the maximum predicted (2573 psia) for FSAR events (see Table 3-19, tests 70, 63, ~nd 64).

To address transition and liquid conditions resulting from FSAR events in these units, one test of each type was performed. The transition test (see Table 3-19, test 69) was performed at a pressure of 2545 psia with 648°F water following the steam init1a11y at the valve inlet. The liquid test (see Table 3~19~ test 68) was

performed with an opening pressure of 2545 psia and a temperature of 647°F. For both tests, the liquid temperatures tested {647 and 648°F) are representative of those expected (654 to 658°F). The opening pressure tested {2545 psia) is slightly below the maximum predicted valve (2575 psia}o Howeverp the analysis asslJlled that the PORVs were not operable and that the safety valve opened 75 psia above its set paint. The tested pressures are considered representative of the peak pressure that would have been predicted had operation of the relief valves been taken into account in the analysis ..

PORVs in Westinghouse units have opening set points of 2350 psia during power operation (reference 3, Table 3-2). The nominal PORV blo\ldown setting in Westinghouse units is 20 psi (reference 31 page 2°7). Therefores the closure pressure for.relief valves in Westinghouse units is 2330 psia. During all tests discussed above, the signa.1 to close the Copes-Vulcan {17 .. 4PH Plug and Cage) relief valve was given at inlet pressures exceeding this value except test 70, the water seal simulation test. During test 70 9 the signal to close was given at an inlet pressure of 2293 psia which is only 1.6 percent below the expected value {2330 psi a) •

Based on the above, the inlet fluid conditions (stateD pressure, and temperature) under which the Copes-Vulcan (17-4PH Plug and Cage) relief valve was tested are considered adequate to represent those expected for FSAR events in the PWR units listed in Table 3°200


The maximum quasi~steady bending moment induced while the Copes~Vulcan (17~4PH Plug and Cage) relief valve was in the opening or closing process during a 11 preload11 test was 43 9 000 in-lb (see Table 3-19, test 64-CV .. p4 .. 2S).

3-43

•

•


As discussed in Section 2.2.4, back pressure is not considered a primary parameter ~affecting the operation of the air operated PORVs tested. Therefore, although ~ack pressures in the range of those expected in the units listed in Table 3-20

were tested, plant-specific justification that the tested values exceed those expected is not required.

3. 7 .2 EXTENDED HIGH PRESSURE INJECTION EVENTS


Extended High Pressure Injection events may result in steam followed by water discharge through the relief valves in some of the Westinghouse units listed .in Table 3-20 (reference 3, Table 5-3). Liquid discharges at the PORV opening set point (2350 psia) with temperatures in the range of 498 to 569°F are predicted (reference 3, Table 5-3).

To represent the conditions resulting from such events, one (1) transition and two (2) 1 iquid tests were performed. The 1 iquid tests (see Table 3-19, Test"s 65 and 68) were performed at opening pressures exceeding 2350 psia and temperatures ranging from 455 to 647°F. As di~cussed in Section 3.7.1, a transition test (see Table 3-19, test 69) was also performed with an opening pressure above 2350 psia and a liquid temperature of 648°F. Closure pressures for all tests discussed above exceeded that expected in the units listed in Table 3-20 with the exception of test 65-CV-174-4W. The closure pressure for this test (2180 psia) was within 6.5 percent of the maximum expected closure.pressure (2330 psia).

Based on the above, the inlet fluid conditions (state, pressure, and .temperature) under which the Copes-Vulcan (17-4PH .Plug and Cage) re lie~ valve was tested, are considered adequate to represent those expected for Extended High Pressure Injection events in the units listed in Table 3~20.


See Section 3.7.1.

3-44


See Section 3.7.l.

3.7.3 COLO OVERPRESSURIZATION EVENTS

The range of expected fluid conditions for Cold Overpressurization events was not provided via the Westinghouse Plant Conditions Justification Report (reference 3) for those units listed in Table 3 .. 20 {see section 1.3 for discussion). TheFefore, justification that such conditions were addressed in the testing of the Copes-Vul can (17-4PH Plug and Cage) relief valve is not presented in this report and will be included as part of the plant-specific evaluations.

3-45

•

w I

.f::> O'I

(PRl/ PORV TEST OATA (])

TMILE 3 • 18 •As TESTED" HARSHALL TEST MATRIX FOR THE COPES-VULCAN (17-4PH PLUG AND CAGE) RELIEF VALVE

CONOITIOHS AT VALVE OPENING TRANSIENT CONDITIONS

VALVE INLET IN ACCUMULATOR oulHToN ~ALY~ INLP RES • WH N nm:oo:n> TEST TEST FLUID PRESS. TEMP. FLulD PRESS. TEMP. (SEC) SIGNAL GIVEN PIPE PRESS.

NO. TYPE (PSIA) (Of) (PSIA) (Of) TO CLOSE VALVE (PSIA) (PSIA)

STEAM STEAM 2455 (SAT.) SAHE AS VALVE INlET 11 2145 595

2 STEAM STEAH 2455 (SAT.) 32 2175 605

3 STEAM STEAM 2430 (SAT.) 3J 2185 615

4 STEAK STEAM 2475 (SAT.) 37 2210 615

5 STE AH STEAH 2475 (SAT.) 30 2195 615

6 STEAM STE AH 2445 (SAT.) 71 2135 195

1 STEAM STE AH 2435 (SAT.) 38 2165 195

8 STEAM STE AH 2445 (SAT.) 38 2195 195

!I STEAM STEAM 2505 (SAT.) 45 2!95 195

10 STE AH STEAM IM45 (SAT.) 46 2175 195

II STEAM STEAM 2455 (SAT.) (2) 2185 615

NOTES: (1) Maximum Quast steady discharge pipe pressure. (2) Not recorded. ( 3) Test results ore for ev11luatton tests only. Total of 22 supplementary valve actuatton cycles 11ere performed under similar conditions.

(EPRl!MYLE PORV TEST DAiA)

iAllU l - !9 "AS USltrn" WlfU rl'llASE 616 if~i wmm1 lfOR JUE

COA'IESi 'llULCM! U-4 I'll PUIG AND f~Gf. 81fUEIF VALVE

HHliRAl CONDBJIOHS(i) TRAHS~fNT CONDlllONS

Ila Ille Dlliiel

At 'llihe Brrah!t ho Accumuhtor Pressure HHftllllllllJ HHl•llll Hu IAllllll ill OB~cb1rge Bending Valve

lei!it Closure Pipe iltonlent Acceleratlon "Kemp il'lreH. Temp l?reu. Duration Signal Press.(2) lnduced(l)lnduced

lest No. Vest iype n11M (Of) (psG11) !FD!! 8d (Of) (psi11» (secondst (psGa) (psia) (An-lb) (g's)

6l-Cll-H4-BS Steal9l $team 602 2.145 Sle4111 676 2.145 6 2,165 450 Bf/A 7.2

64-ll:ll-174-25 StHllll Stea11! 602 2.145 Steillll 677 2,145 6 2,365 410 4l,OOO 6.9 w «Preioi!!d» 8

.g:. '8

65-CV- 174-418 Nater' Malter 455 2,535 lo!1lerr 452 2.535 4l 2,180 425 ~IA 5.9

66-Cll-U4-llt! !ili!l~eir Mater 4<12 us MDter 441 615 ftO t!ilO B!U NIA 1.8

61-CIJ-174-51-8 Wat ell' Maler IOfi> '575 Mateir ID!! . us no 512 8.8 NIA 1.1

60-CV-B74-6W w~teir Mater 647 2,545 &iaierr 651 :£,545 5 2,340 5l1 ~IA l.l

6!!-CV-Wl-15/M "8'r&ns6Uo111 Steam Ui 2.545 Maler 648 2.M!ii 5 2.112 5416 MIA 2.6

70-Cll- ft/'4-0U/M '6aaer Maler 115 Zo115 Malter 651 . 2.115 B6 2,291 618 MIA 12.2 Se&1D $9mu8at hm

Ii) GN2 ~ORV Actuation UDYage pressure for aHI tests Mas 86 ! R ~s6g ------~ 2 _flo btick filiressuire illrBfh:e r.?as 11aselfi 6111 the Co11es-lluiolll (U-4) IJ>Oltlf ~esth19 1l Va811e :>hmm corresponds to lite maximum moment a111,Dle!D llihlle Ute 1111lve was In tile t11pei&hag/dosing process

•·

•

Table 3-20

PWR UNITS WITH COPES-VULCAN (17-4PH PLUG AND CAGE) RELIEF VALVES

Unit NSSS Vendor Comments

H. 8. Robinson 2 (CPL) Westinghouse Three loop unit

Turkey Point 3, 4 (FPL, FLA) Westinghouse Three loop units

Prairie Island 1, 2 (NSP, NRP) Westinghouse Two loop units

Sal em 1, 2 {PSE, PNJ) Westinghouse Four loop units

Surry 1, 2 (VPA,. VIR) Westinghouse Three loop units

Point Beach 1, 2 (WEP, WIS) Westinghouse Two loop units

Note: Justification of conditions resulting from Cold Overpressurization events is not presented for the units shown above. Such justification will be provided as part of these utilities' plant-specific ev~luations •

3-48

3.8 COPES-VULCAN (316 W/STELLITE ·PLUG AND 17-4PH CAGE)

Tables 3-21 and 3-22 present the conditions under which the Copes-Vulcan (316

•

w/Stellite Plug and 17-4PH Cage) relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-23 lists the PWR units using or planning to use this relief valve design. The following sections

•

provide specific justification that the inlet fluid conditions tested are representative of those expected for FSAR, Extended High Pressure Injection, and Cold Overpressurization* events in these units. In addition, discharge piping effects on valve operability (back pressure and bending moment) imposed on the CopesVulcan (316 w/Ste11ite Plug and 17-4PH Cage) relief valve are discussed.

As shown in Table 3-23, Copes-Vulcan (316 w/Stellite Plug and 17-4PH Cage) relief valves are currently utilized only in Westinghouse 2, 3, and 4 loop units. The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.8 .• 1 FSAR EVENTS


FSAR events can result in steam, transition, and water conditions at the inlets of relief valves in some of the units listed in Table 3-23 (reference 3, Tables 5-1 and 5-2**). In addition, some of these units utilize a cold water seal design.

As discussed in Section 3.1, an assl.ITled opening time in the range of two (2) seconds for the Copes-Vulcan relief valve, necessitates demonstration of operability at inlet pressures up to the maximum predicted during FSAR Overpressurization events. The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in Westinghouse units, is 2573 psia (reference 3, Table 5-1). The maximum pressure expected during FSAR events resulting in transition or liquid discharge, in the units listed in Table 3-23, is 2575 psia (reference 3,

*Cold Overpre~surization conditions justified only for plants as noted in Table 3-23. Such justification for the other units listed will be provided as part of these utilities• plant-specific evaluation.

**Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses performed to·obtain these conditions assumed that the PORVs were not operable. However, if the PORVs are operable conditions representative of those presented can be expected at the PORV inlets as well •

3-49

Table 5-2). From the same table, the range of expected liquid temperatures for these units is defined as 568.7 to 672°F.

One (1) water seal siRnJlation and one (1) steam test were performed on this relief valve design with inlet opening pressures (2715 and 2725 psia) well in excess of the maximum predicted (2573 psia) for FSAR events (see Table 3-22, tests 78 and 71) 0

To address transition and liquid conditions resulting from.FSAR events in these units, one (1) transition and twd (2) liquid tests were performed. The transition test (see Table 3-22, test 77) was performed at a pressure of 2532 psia with 657°F

·water following the steam initially at the valve inlete The liquid tests (see Table 3-22, Tests 72 and 76) were performed with opening pressures of 2545 and 2535 psia and a temperature of 446 and 647°F, respectively. Far these tests, the liquid temperatures tested (446 to 647°F) were representative of those expected (569 to 672°F). The opening pressures attained during the transition and liquid tests (2545, 2535, and 2532 psia) were slightly below the maximum predicted value (2575 psia). However, the analysis asslllled that the PORVs were not operable and that the safety valve opened 75 psia above its set point. The tested pressures are considered representative of the peak pressure that would have been predicted had operation of the relief valves been taken into account in the analysise

PORVs in Westinghouse units have opening set points of 2350 psia during power operation (reference 3, Table 3-2)o The nominal PORV blmiidown setting in Westinghouse units is 20 psi (reference 3» page 2°7)e Therefore, the closure pressure for relief valves in Westinghouse units 1s 2330 psia. During all tests discuss~d above, tfle signal to close the Copes-Vulcan relief valve was given at i.nlet pressures exceeding this value except during test 72, a ful 1 pressure, 446°F water testo During this test, the closure ~ignal was given at an inlet pressure of 2170 psia which is within 6c9 percent of the in-plant value (2330 psia}o

Based on the above 9 the inlet fluid conditions {state, pressure, and temperatui'"e) under which the Copes-Vulcan (316 w/Ste11ite Plug and l7m4PH Cage) relief valve was tested are consid~red adequate to represent those expected for FSAR events in the PWR units listed in Table 3-230

3-50

•

•


No specific "prel oad" test was perfonned on the Copes-Vulcan (316 w/Stell ite Plug .• and 17-4PH Cage) relief valve since the Copes-Vulcan valve's ability to operate

under a bending load was demonstrated while testing .this valve with 17-4PH internals (see Table 3-19, test 64-CV-174-2S).


As discussed in Section 2.2.4, back pressure is not considered a primary parameter affecting the operation of the air operated PORVs tested. Therefore, although back pressures in the range of those expected in the units listed in Table 3-23 were tested, plant-specific justification that the tested values exceed those expected is not required.



Extended High Pressure Injection events result in steam followed by water discharge through the relief valves in some of the Westinghouse unit.s listed 1n Table 3-23 (reference 3, Section 5-3). Liquid discharges at the PORV opening set point (2350 psia) with temperatures in the range of 498 to 569°F are predicted (reference 3, Table 5-3).

To represent the conditions resulting from such events, one (1) transition and two (2) liquid tests were perfonned. The liquid tests (see Table 3-22, Tests 72 and 76) were performed at opening pressures exceeding 2350 psia and temperatures· ranging from 446 to 647°F. As discussed in Section 3.8.1, a transition test (see Table 3-22, test 77) was also performed with an opening pressure above 2350 psia and a liquid temperature of 657°F.

Closure pressures for all tests discussed above exceeded that expected in the units listed in Table 3-23 with the exception of test 72-CV-316-3W. As discussed in Section 3.8.1 the closure pressure for this test was within 6.9 percent of the maximum expected closing pressure (2330 psia).

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Copes-Vulcan (316 w/Stellite Plug and 17-4PH Cage) relief valve was tested are considered adequate to represent those expected for Extended High

• Pl'essure Injection events 1n the units listed in Table 3-23.

3-51

BENDING t>tlMENT INDUCED

See Section 3.8.1.


See Section 3.8.1.

3 .8 .3 COLD OVERPRESSURIZATION EVENTS

The range of expected fluid conditions for Cold Overpressurization events was provided via the Westinghouse P1ant Conditions Justification Report (reference 3) only for those units as noted in Table 3-23 (see section 1.3 for discussion). Therefore, justification that such conditions were addressed in the testing of the Copes.,Vulcan (316 w/Stellite Plug and 17-4PH cage) relief valve is provided only for those units so designated. Such justification for the other units utilizing this valve design will be included in the plant-specific evaluations.

INLET FLUID SiAiE/PRESSURE/TEMPERATURE

The full range of potential Cold Overpressurization conditions for all Westinghouse units eva1u·ated is depicted in Figure 5-1 of reference 3. As discussed in Section 5.4 of reference 3, steam, transition, and subcooled.water conditions occurring over a large range of pressure and temperature are possible during such events in the Westinghouse units evaluated. From Figure 5-1, ~ference 3, ft 1s. noted that fu11 pressure .(2350 psi a) liquid discharges are expected over a temperature range of 250°F to 650°F. Based on this figure and dhcuss1ons in Section 5.4 of reference 3, lower pressure (300 psi al 11 quid discharges are expected aver a liquid temperature range of 100 to 400°F.

The full pressure (greater than 2350 psia) steam and transition tests previously discus$ed (Table 3-22 1 tests 71 and 77) envelop expected steam and transition conditions resulting from such events.

liquid tests over a range of pressures and temperatures representative of those shown in Figure 5~1, reference 3, were also performed. Full pressure (greater than 2350 psh.) tests with liquid temperatures ranging from 274 to 647°F were performed (see Table 3-22. tests 75 9 72w and 76). In add1tioni reduced pressure (....075 psia) tests with liquid temperatures ranging from 105 to 442°F were performed (see Table 3-22, tests 73 and 74).

3-52

•

•

The valve closure pressures for one of the full pressure liquid tests (test 76) discussed above exceeded the maximum expected closure pressure (2330 psia). The

· closure pressures for the other tests (tests 72 and 75) were 2170 and 2230 psi a • which are within 6.9 percent of the maximum expected value (2330 psi a).

•

Valve closure pressures for all reduced pressure tests were well above the minimum expected opening pressure (300 psia) corresponding to the lower end of the expected liquid temperature range (100-400°F).

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Copes-Vulcan (316 w/Stellite Plug and Cage 17-4PH Cage) relief valve was tested are considered adequate to represent those expected for Cold Overpressurization events in the units for which such conditions were provided by the Westinghouse Pl ant Condition Justification Report (see Table 3-23).


See Section 3.8.1.


See Section 3.8.1 •

3-53

-------·-

w J. ~

tflPRI/ MllSMU. PORV TUT DATA (J)

VMll J - 2B

"'AS iHVIEO" MRSllAU. TESV MflUI !FOil 11tE COHS-111.CM 016 11/iiTUUTIE IP&.US Alllll 11·4Pll CAGE) !IEUff VALVE

CONDITIONS AV WAlW~ ~P!RIMG 'l!UllSIUT COllDlllONS

~Al.VIE BHUY . BM ACCUHU!..ATOll oo11n. HHl.11iH lttUllll11 TEST HST RUHi PRESS. TflF. FLUID PRESS. UMP. 4SEC) SIGHAL GIVEN PIP! l'llESS. HO. il!Pf: (l'SRA) C0f) (PSIA) (Of) 10 CLOS[ VALVE (PSIAI

PSIA

i ~TOH STEAM 1471 «SAT.) SW AS WALY£ INLET 41 1155 lil5

2 SIEA!t SifM 2460 (SAT.B 18 2155 IJ5 J SHAH SJ(M 2450 (SAT.) »• 2155 Ill 4 STEM STEM 2455 (SAT.) ., 2165 115

--- .I SiEAH STEM 2465 fSAi.) zo 2165 5J5

i HIEAH SUM !460 (SAT., 15 Hl5 115 'fJ SUAH STEM HJS «sav.p II 2155 115

I STEM STEM 2450 (SAJ.) 11 2155 Ill !II $UM PEAM !455 (SAT.) 2:ll 2165 115

10 SiEAH STUM 2465 (SAV.) 11 1115 115

n STEM 5HM . !460 QSAi.) co 21Ci0 il!i

C\!OUS:

lit llui- QlaHS 11t11acB.11 d811~illll1'11& 11>lfllft pr1Hs11re. Z Not 11'ecorcBecB. J iest results 1r1 for owa1ua&~om Rests Qll1J. io&8~ Qf 86 supp1le!llentQr» v11ve 1ctu1t8on gmstA ll!Qr& 111trforwed under 1t•l11r 'ondftt!on1 •

•

• • (EPRl/WYl.E PORV TEST DATA)

TABLE 3 - ZZ •AS TESTED• WYLE PHASE Ill TEST MATRIX FOR THE

COPES VULCAN ll6 W/STELL ITE PLUG AUD 11-4 PH CAGE RELIEF VALVE

IHIT IAL_ CO~ITl!!!!~t!L ·······-·----·-- TR~NSIEf!T CONDl.llQ.N_S ___

Valve Inlet

At Valve Inlet In Accumula~Qr ____ Pressure Haxtmum Max l111um Haxllrium al Discharge Bending Valve

Test Closure Pipe Moment Acceleration Temp Press. Temp Press. Duration Slgnal Press.(Z) Induced Induced

Test No. 1~!!..!tf!. F1uld («!_l ie.~!!!1 Flu Id (~J (P.~ ~~I ~ecl!.ndsl (p~~~L.. (p!!!L_ (In:!~) . (f_~J_ ___ --------·

71-CV-316-lS Steam Steam 6GZ 2.715 Steam 617 Z.7l!i 6 Z.lJl 460 ff/A 6.6

12-CV-316-JW Water Water 446 2.545 Water 442 2.545 4 z.110 307 N/A 4.6

73-CV-316-411 Water Water 44Z 675 Water 44Z 675 10 6Z6 ~9] N/A 1.1 w I U1 <n 74-CV-ll6-5W Waler Waleli' 105 675 Water 9Z 675 10 562 174 ff/A 3.6

75-CV-316-6W Water Water Z74 2.110 Water 261 2 .110 5 2.230 (3) ff/A 7.6

76-CV-316-2W Water Water 647 z.535 Water 654 2.535 5 2.350 531 N/A 3.9

77-CV-316-7S/W Transition Ste4111 670 2.532 Water 657 2.532 6 2.343 555 N/A 12.6

78-CV-316-0W/W Water Water 134 2.725 Water 654 2.12s 16 2.350 640 N/A 7.0 Seal Stmulatlon

79-CV-ll6-9N/W Transit ton GHz 262 l.533 Water 299 1.533 6 I.JOO 150 N/A 5.1

11> GHz PORV Actuation UHage Pressure for. a.II Tests was 86 (!. 1) pstg. · 2) No Back Pressure Orifice was used fo the Copes-Vulcan 316 PORV Testing. (3) PS-4 In operative during this test.

Table 3-23

PWR UNITS WITH COPES-VULCAN (316 W/STELLITE PLUG ANO 17-4PH CAGE) RELIEF VALVES

.!:!!!i!. NSSS Vendor ColRnents

Joseph M. Farley 1, 2 (ALA 1 APR)3 Westinghouse Three loop units

Shearon Harris 1, 2, 3,4 (CQL, CRL, CSL, CTL)3 Westinghouse 'Three loop units

Byron l, 2 (CAE, CSE) Westinghouse Four loop units

Braidwood l, 2, (CCE, COE) Westinghouse Four loop units

Connecticut Yankeel Westinghouse Two loop unit

Indian Point 2 (IPP)2,3 Westinghouse Four loop unit

South Texas l, 2 (TGX, THX) Westinghouse Faur loop units

Trojan (POR)3 Westinghouse Four loop unit

Indian Point 3 (INT)2,3 Westinghouse Four loop unit

Marble Hill 1, 2 (PBJ, PCJ)3 Westinghouse Four loop units

Seabrook l» 2 (NAH, NCH)3 Westinghouse Four loop units

Ginna (RGE)3 . Westinghouse Two loop unit

Virgil Ca Summer (C~E)3 Westinghouse Three loop unit

Sequoyah 2 (TEN) Westinghouse Four 1 oop unit

Commanche Peak l, 2 (TBX, TCX} Westinghouse Four loop units

Zion 1, 2 (CWE, COM)j Westinghouse Four loop units

lNo specific test condition justification presented for this unite Such justifiQ cation will be provided as part of this utility's plant~specific evaluation.

2rhe relief valves in these units utilize a Haynes Noc 25 instead of a l7e4PH cag~.

3Justification for conditions resulting, from Cold Overpressurization events is not presented for these units. Such justification will be provided as part of these utilities• plant-specific evaluations.

I

3-56

•

•

•

•

•

. 3 .9 MUESCO CONTROLS

Tables 3-24 and 3-25 present the conditions under which the MUESCO Controls relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-26 lists the PWR unit utilizing the MUESCO Controls relief valve design. The following sections provide specific justification that the inlet fluid conditions tested are representative of .those expected for FSAR and Extended High Pressure Injection events in this unit. In addition, discharge piping effects on valve operability (back pressure and bending moment) imposed on the MUESCO Controls relief valve are discussed.

As shown in Table 3-26, MUESCO Controls relief valves are currently utilized only in San Onofre 1, a Westinghouse 3 loop unit. The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.9.1 FSAR EVENTS


FSAR events result fo steam conditions at the inlets of relief valves in the San Onofre 1 unit (reference 3, Tables 5-1 and 5-2*). In addition, this unit does not utilize a water seal design •

As discussed in Section 3.1, an assumed opening time in the range of two (2) seconds for the MUESCO Controls relief valve, necessitates demonstration of operability at inlet pressures up to the maximum predicted during FSAR Overpressurization events. The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in 3-loop Westinghouse units, is 2555 psia (reference 3, Table 5-1).

One (1) steam test was performed on the MUESCO Cont~ols relief valve with an inlet· opening pressure (2755 psia) well in excess of the maximum predicted (2555 psia) for FSAR events (see Tab 1 e 3-25, test 80) •

PORVs in Westinghouse units have opening set points of 2350 psia during power operation (reference 3, Table 3-2). The nominal PORV blowdown setting in

*Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses performed to obtain these conditions assumed that the PORVs were not operable. However, if the PORVs--are operable conditions representative of those presented can be expected at the PORV inlets as well.

3-57

Westinghouse units is 20 psi (reference 3, page 2-7). Therefore, the closure pressure for relief valves in Westinghouse units is 2330 psia. During the test discussed above, the signal to close the MUESCO Controls relief valve was given at an inlet pressure (2500 psia) exceeding this value.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the MUESCO Controls relief valve was tested aFe eonsidered adequate to represent those expected for FSAR events in the San Onofre l unita


The maximum quasi-steady bending moment induced while the MUESCO Controls valve was in the opening or closing process during a 11preload" test was 24,000 in .. lb

(see Table 3-255 test 81-MU~2S}.


As discussed in Section 2.2a4, back pressure is not considered a primary parameter affecting the operation of the air operated PORVs tested~ Therefore, although back pressures in the range of those expected in the San Onofre l unit were tested plant-specific justification that the tested values exceed those expected is not requireda

3o 9e2 EXTENDED HIGH PRESSURE INJECTION EVENTS


Extended High Pressure Injection events result fn steam followed by water discharge through the relief' valve in Westinghouse 3 .. 1 oop units (reference 3, Section 5-3)a Lfquid discharges at the PORV opening set point (2350 .psia)-With temperatures in the range of 498 ·to 502°F are predicted (reference 39 Table 5-3).

' , .

To Fepresent the conditions resulting from such events, one transition and two (2) liquid tests were perfonned. The liquid tests (see Table 3m259 tests 82 and 85) were perfonned at opening pressures exceeding 2350 psia and tempeFatures ranging from 455 to 645°F. A transition test (see Table 3-251 test 86) was also performed with an opening pressure above 2350 psia and a liquid temperature of 652°F.

Closure pressures for all tests discussed above exceeded that expected in the San Onofre 1 unit (2330 psia).

•

•

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the MUESCO Controls relief valve was tested are considered adequate to represent those expected for Extended High Pressure Injection events in the San

~ Onofre 1 unit.


See Section 3.9.1.


See Section 3.9.1.


The range of expected fluid conditions for Cold Overpressurization events was not provided via the Westinghouse Pl ant Conditions Justification Report (reference 3) for the San Onofre 1 unit {see Section 1.3 for discussion). Therefore, justification that such conditions were addressed in the testing of the M1JESCO Controls relief valve are not presented herein and will be included in their plant-specific, evaluation.

3-59

w I 0\ 0

•

lPAll IMliSMlll. POllV 1m DA1A « J)

VMi.lt ll - .H

"'A! THTl!:D" IMISW.&. VHT Mlllll fllA VII£ lllESCG UUIF 'AU£

COHDIVIOMS Ai WAtVi OPENVIG . llM.Vf BNLH 819 ACCIHILAJOR oollHOlll UST UH fllllC PUSS. mt.I'. flDlo PRESS. n ... (SEC) NO. iV!li (PSIA) (Of) (PSIA) (Of)

I ii YEAH STEM 1435 (SAT.) iH AS 11Al VE 1111.ET 151 z STEAM SlEM 242$ «SAi.» Di J SUM STEAM 2415 (SAV.B !Ii 4 SHAH 5TEMI Z4H (SAT.) !O

i SUAM SUM IUS (SAY.i U!

' $i(Al'i STEAltl 2U5 CSAi.b H

i SU:AH SHM !455 (SA"i_D RI

• SUNi SUM !455 (SAY.» D

' SUM STEM 2455 (SAY.) 11 10 STEM SUM 2«45 (SAT.B I& 118 SiEAH iUAH !~5$ (SAV.~ «O

llOTE5: h) lllxl- QuH9 1tea111!1 ifl8acl!u11rgfl 98gH1 pn!nura. (2) Hot irecorrdecl.

V•AHSIEIJ COHOIJIONS

mn.1llhH r.nw0 UGHAL GIVEN PIPE PllESS. TO CLOS[ 11ALVf (PSIA)

PSIA

2385 HS

2375 255

1375 155

2J75 255

2l75 155

2421 80

2415 35

2U5 15

2405 15

2400 15

2405 15

U) 1e1t ire11u'it11 11r@ for ev11111tOdlll le1t1 on1y. Total of 14 supp1-nt1r1 .111he 1ctu1tlon crcllll! !Hll'fl perrf11~cl 11nderr &l•l1tr c:ondlUons.

• •

w I 0\ .....

(EPRl/WYLE PORV TEST DATA) TABLE 3 - 25

•AS TESTED" WYLE PHASE 111 TEST MATRIX FOR THE HUESCO CONIROLS RELIEF VALVE

INITIAL CONDITIONS(l)

Valve Inlet

At Valve Inlet 11'!_8fcum_u1ator Pressure at

Test Closure Temp Press. Temp Ptress. Duration Signal

Test No. J.est Type Fluid ~fl J~!_la) fluid j~u_ .(es la) _(seconds) ( ps ~.!L ·----80-ttJ-lS Steam Steam 683 2,755 Steam 678 2.755 12

8l-MU-2S Steam Steam 670 2,,535 Steam 665 2.535 6 (Pre load)

82-HU-3W Water Water 455 2,536 Water 453 20sl6 9

8l-HU~4W Water Water 449 674 Water 452 674 11

84-Hll-5W Water Water 106 617 Water 99 677 11

tl5-HU-6W Water Water 645 2,534 Water 651 2,,5]4 12

86-HU-7S/W Transftton Steam 670 2,540 Water 652 2,540· 12

(1) Glf2 PORV Actuation U11age pressure for all tests was 49 .!. 1 pslg. (2) No back pressure orlf Ice was used fn the Muesco PORV testing. (3) Values shown corresponds to the maximum mmnent Induced while valve ts In the opening/closing process.

2,500

2,418

2,410

674

649

2,442

2,440

••

IBOOi!ENL COlf!!lW!.tf~-----·-

HaxlmlllA Maximum Maximum Discharge Bending Valve Pipe Moment Acceleration Press.(2) lnduced(3)1nduced _(psla) (In-lb) (~~s)

550 N/A 5.1

50 24,000 5.0

184 N/A 9.6

78 N/A 1.1

84 N/A 20.2

110 N/A 4.4

106 N/A 4.3

Unf t

San Onofn 1 ( SCE) 1

Table 3-26

PWR UNITS WITH MUESCO CONTROLS RELIEF VALVES

NSSS Vendor

Westinghouse

Comments

Three 1 oop uni t

lJustiffcation for c~ndftions resulting from Cold OYerpressurization evefits is not presented for this unit. Such justification wi11 be provided as part of this utility's Dlant-specific ev~1uat1ono

3.10 FISHER CONTROLS

Tab 1 es 3-27 and 3-28 present the conditions under which the Fi sher Contro 1 s relief valve design was tested during the Marshall and Wyle Phase III test programs. respectively. Table 3-29 lists the PWR units using or planning to use the Fisher Controls relief valve design. The following sections provide specific justification that the fnlet fluid conditions tested are representative of those expected for FSAR, Extended High Pressure Injection, and Cold Overpressurization* events in these units. In addition, discharge piping effects on valve operability (back pressure and bending moment) imposed on the Fisher Controls relief valve are discussed.

I

As shown in· Table 3-29, Fisher Controls relief valves are currently utilized only in Westinghouse 3 and 4 loop units. The expected range of inlet fluid conditions for Westinghouse units is presented in reference 3.

3.10.1 FSAR EVENTS


FSAR events can result fn steam, transition, and water conditions at the inlets of relief valve~ in some of the units listed in Table 3-29 (reference 3, Tables 5-1 and 5-2**). In addition, all these units utilize a water seal design.

As discussed in Section 3.1, an assumed opening time in the range of two (2) seconds for the Fisher Controls relief valve, necessitates demonstration of operability at inl~t pressures up to. the maximum predicted during FSAR Overpressurization events. The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in Westinghouse 3 and 4 loop units, is 2555 psia (reference 3, Table 5-1). The maximum pressure expected during FSAR events resulting in transition or liquid discharge, in the units listed in Table 3-29, is 2575 psia. (reference 3, Table 5-2). From the same table, the range of expected liquid temperatures for these units is defined as 553.8 to 636.6°F.

*Cold Overpressurization conditions justified only for plants as noted in Table 3-29. Such justification for the other units listed will be provided as part of their plant-specific evaluations.

**Table 5-2 provides expected conditions at the inlet of pressurizer safety valves. The NSSS vendor analyses performed to obtain these conditions assumed that the PORVs were not operable. ·However, if the PORVs are operable conditions representative of those presented can be expected at the PORV inlets as well.

3-63

Two (2) water seal simulation and two (2) steam tests were perfonned on the Fisher Controls relief valve with inlet opening pressures (2704, 2700, 2760, and 2760 psia) well in excess of the maximum predicted {2555 psia) for FSAR events {see Table 3-28, Tests 95, 96, 87, and 88). To address transition and liquid conditions resulting from FSAR events in these units, one transition and two liquid tests were perfonned. lhe transition test (see Table 3-28, test 94) was perfonned at a pressure of 2530 psia with 653°F water fallowing the steam initially at the valve inlet. The liquid tests (see Table 3-28, tests 89 and 93) were perfonned at opening pressures exceeding 2350 psia and temperatures ranging from 452 to 648°Fo For these tests the liquid temperatures tested (452, 648 and 653°F) were representative of those expected (553e8 to 636.6°F)e

The opening pressures for the transition and liquid tests {2530, 2664 and 2536 psia) were representative of the maxi!ll.1111 predicted value {2575 psia)o

PORVs in Westinghouse units have opening set points of 2350 psia during power ope~ation (reference 3, Table 3e2). The nominal PORV b10\!ddown setting in Westinghouse units is 20 psi (reference 3, page 2e7). Therefore, the closuFe pressure for relief valves in Westinghouse units is 2330 psi a. During all tests discussed above, the signal to close' the Fisher Controls relief valve was given at inlet pressures exceeding this value.

Based on the above, the inlet fluid conditions (state, pressure8 and temperature) under which the Fisher Controls relief valve was tested are considered adequate to represent those expected for FSAR events in the PWR units listed in Table 3-29.


The maxil1lllll ·quasi-steady bending moment induced while the Fi sher Controls valve was fn the opening or closing process during a 111 preload19 test was 38,300 i-n-lb (see Table 3-28, test 88..f'S-2S).


As discussed in Section 2o2.4s back pressure is not consideFed a primary parameter affecting the operation of the air operated PORV 9 S tested. Therefore» although baek pressures in the range of those expected fn the units listed fn Table 3~29 were tested, plant~specific justification that the tested values exceed those e~pected is not requiredo

3-64

3.10.2 EXTENDED HIGH PRESSURE rNJECTION EVENTS


Extended High Pressure Injection events result in steam followed by water discharge through the relief valves in Westinghouse 3 and 4 loop units (reference 3, .Table 5-3). Liquid discharges at the PORV opening set point (2350 psia) with temperatures in the range of 498 to 569°F are predicted (reference 3, Table 5-3).

To represent the conditions resulting from such events, one transition and two (2) liquid tests were performed. The liquid tests (see Table 3-28, tests 89 and 93) were performed at opening pressures exceeding 2350 psia and temperatures ranging from 452 to 648°F. As discussed in Section 3.10.1, a transition test (see Table 3-28, test 94) was also performed with an opening pressure above 2350 psia and a liquid temperature of 653°F.

Closure pressures for all tests discussed above exceeded that expected (2330 psia) in the units listed in Table 3-29.

Based on the. above, the inlet fluid conditions (state, pressure, and temperature) under which the Fisher Controls relief valve was tested are considered adequate to

• represent those expected for Extended High Pressure Injection events in the units 1 i sted in Table 3-29.


See Section 3.10.l.


See Section 3.10.1.


The range of expected fluid conditions for Cold Overpressurization events was provided via the Westinghouse Plant Conditions Justification Report (reference 3) only for those units as noted in Table 3-29 (see Section 1.3 for discussion). Therefore, justification that such conditions were addressed in the testing of the Fisher Controls relief valve is provided only for those units so designated. Such justification for the other units listed will be included in the plant-specific evaluations.

3-65

INLET FLUID STATE/PRESSURE/TEflf>ERATURE

The full range of potential Cold Overpressurization conditions for all Westinghouse units evaluated are depicted in Figure 5-1 of reference 3. As discussed in Section 5.4 of reference 31 steam, transition, and subcooled water conditions occurring over a large range of pressure and temperature ~re possible during such events in the Westinghouse units evaluatedo

Fran Figure 5-1, reference 3, ft fs noted that full pressure (2350 psia) liquid discharges are expected over a temperature range of 250°F to 650°F. Based on this figure and discussions in Section 5.4 of reference 3, lower pressure (300 psia) liquid discharges are expected over a liquid temperature range of 100 to 400°F.

·TI'ie fu11 pressure (greater ~han 2350 psia) steam and transition tests previously discussed (Table 3 .... 281 tests 87 and 94) envelop expected steam and transition conditions resulting fran such events.

liquid tests over a range of pressures and temperatures representative of those shown in Figure 5-1, reference 3, were also performed. Full pressure {greater than 2350 psia) tests with liquid temperatures ranging from 264 to 648°F were perfonned (see Table 3-28, tests 92 and 93). In addition, reduced pressure (-685 psia) tests with liquid temperatures ranging from 101 to 447°F were per= fonned (see Table 3a28» tests 91 and 90)o

Valve closure pressures for all full pressure liquid tests discussed above exceeded the maximllll expected closure pressure (2330 psi a). Valve closure pressures for all reduced pressure tests were we11 above the expected opening pressure (300 psia) eorresponding to the lower end of the expected liquid temperature range {100-400°F).

Based on the aboves the inl~t fluid conditions {state9 pressure9 and temperature) under whieh the Fisher Controls relief valve was tested 9 are considered adequate to represent those expected for Cold Overpressurization events in the applicable units designated in Table 3-29a

3-66

•

•

•


See Section 3.10.1.


See Section 3.10.1 •

3-67

w 8

°' 00

t:PRI/ MARSHALL PORV TEST DATA (I) T.Mi.!E J - 21

"AS 1Jll:$Vf0m l!WISHAU. i!E$1J /Jtl!TIRn FOi ill( fl SHEii COtfJlllOLS llUlflf lfAUE

~ONDITIONS ii WALllll: @PiNSNE llANSDfli CONDITIONS llAlYf H!LH

IESi 'IJ($1f f!.UID !PAHS. 1TEIP. llG. 1WP£ (PSHA) (Of)

i STfNI STEAl\l 2435 (SAJ.»

2 Sll!'NI SJEM 2420 (SAi,j

J SUM STEAM llll!i (SAT.)

'l SUM SUM· !'lH (SAi'.)

5 5UAll SJEAH 2415 (SAV.)

' STEM STEAi!! 24115 (SAY.i

1 SifAll STEM HlG UAT.)

• STEM STEAM 2415 (SAl.)

!I STEM STEAlll !UC (SAi.)

10 STEM SHAM 2415 (Slli.,

II STEM !HEAii . HJS (SAT.)

1!101U: hb 11 .. 1 .... Quu8 atiHci1 d8Ach11r110 {l)IPfl Qlnt511111re. (2) !lot l!'CCOl!'ded.

BN ACtJ,llUlAlOll ouilU1111 JA!U. 1llhll ifLDlo Pam. nhP. (SIEC) SiGHAl. GIVEN

(PSUI) (Of) TO CLOSE VALVE PSIA

SAHE AS ~Ai.ti( iHlfV ea 22l5

J9 22l5

" 2250

•• 2255

Ji 2255

712 22l0

IO 2265

14 2255

rJ 2255

22 2255

bl 2275

IU.'l'J&P1 llPE PRESS.

(!'SIA)

no .. ~ 4115

4115

4!15

155

B55

155

155

160

4115

(J) le5t resul&B are Wor •~~illllafion gez~g only. loi1i of 28 uu~pie11eni8r1v1ive1ctu1tlon cycles 1111rs pmirfofllllci 1111der 1l•ll1r conciltlong.

···----·

• (EPRl/WYLE PORV TEST DATA) TABLE 3 - 28

"AS TESTED" WYLE PHASE Ill nsr MATRIX FOR THE FISHER CONTROLS RELIEF VAUE

INITIAL CONDITIONS(I) TRAN~!~~Lf!!_NDITIONS ____

Valve Inlet

At Valve ln!!l ___ -·--· .. _! ~_J\cc~mu l_ato!'.'._ __ Pressure Maximum Maximum Maximum at Discharge Bending Valve

Test Closure Pipe Holllent Acee lerallon Temp Press. Temp f'ress. Duration· Signal Press.(2) lnduced(3)1nduced

J~t No. Test~ .f!.!!J~ l~fl .(P~~J f.luld ffl j~ (seconds) (P.~!!L (j!~laL_ (~_!!-lbL_ (9's)

87-FS-IS Stea1n Steam 683 2,760 Steam 678 2',760 6 2,395 370 N/A 2.1

80-FS-2S Steam Steam 683 2,760 Steam (Pre load)

678 2,760 6 2,400 330 38,300 2.0

09-fS-311 Water Water 452 2,664 Water 454 2,664 5 2,430 350 ff/A 2.1 w I 0\ 90-FS-4W Water Water 447 685 Water 456 605 10 666 138 N/A I. 7 ID

91-FS-SW Waler Water 101 684 Water 94 684 10 612 N/A 1.8

92-FS-6W Water Water 264 2,668 Water 257 2,668 4 2,390 410 N/A 3.2

93-FS-7W Water Water 648 2,536 Water 650 2,536 6 2,395 400 N/A 2.5

94-FS-OS/W Transition Steam 669 2,530 Water 653 2,530 6 2,380 312 N/A 7.4

95-FS-911/W Water Water Seal

203 2,704 Water 655 2,704 15 2,420 450 N/A 19.5

Simulation

96-FS-IOW/W Water Water 123 2,700 Water 657 2.100 15 2,400 450 ff/A 11.2 Seal Slmulallon

(1) GN2 PORV Actuation Ullage pressure for all tests was 69 (!.) 1 pslg. (2) No back pressure orlf tce was used in the Huesco PORV testing. (3) Values shown corresponds to the maximum moment Induced while valve Is In the opening/closing process.

Table 3-29

PWR UNITS WITH FISHER CONTROLS RELIEF VALVES

.Y!!.!l NSSS Vendors

Beaver Valley 2 (DMW)l Westinghouse

Watts Bar 1, 2 (WATP WBT). Westinghouse

Co11111ents

lhf"ee 1 oop unit

Faur 1 oop units

!Justification of conditions resulting from Cold Overpressurization events is not included for this unit. Such justification will be provided as part of this Utility's plant-specific evaluation.

3-70

•

3.11 GARRETI

Tables 3-30 and 3-31 present the conditions under which the Garrett relief valve design was tested during the Marshall and Wyle Phase III test programs, respectively. Table 3-32 lists the PWR units using or planning to use the Garrett relief valve design. The following sections provide specific justification that the inlet fluid conditions tested are representative of those expected for FSAR, Extended High Pressure Injection, and Cold Overpressurization* events in these units. In addition, discharge piping effects on valve operability (back pressure and bending moment) imposed on the Garrett relief valve are discussed.

As shown in Table 3-32, Garrett relief valves are currently utilized only in Combustion Engineering and Westinghouse 4 loop units. The expected range of inlet fluid conditions for Combustion Engineering and Westinghouse units are presented in references 2 and 3, respectively.

3.11.1 FSAR EVENTS


FSAR events can result in steam, transition, and water conditions at the inlets of relief valves in the Westinghouse units (reference 3, Tables 5-1 and 5-2**) and steam conditions in the CE units (reference 2, Table 5-19) listed in Table 3-32. In addition, some of these units utilize a cold water seal design.

As discussed in Section 3.1, as assumed opening time in the range of two (2) seconds for the Garrett relief valve, necessitates demonstration of operability at inlet pressures up to the maximum predicted during FSAR Overpressurization events. The maximum expected PORV inlet pressure for FSAR events resulting in steam discharge in the applicable Combustion Engineering and Westinghouse 4 loop units, is 2752 psia (reference 2, Table 5-19). The maximum pressure expected during FSAR events resulting in transition or liquid discharge, in the units

*Cold Overpressurization conditions justified only for plants as noted in Table 3-32. Such justification will be provided for the other units listed as part of their plant-specific evaluations.

**Table 5-2 provides expected conditions at the inlet of pressurized safety valves. The NSSS vendor analyses performed to obtain these conditions assumed that the PORVs were not operable. However, if the PORVs are operable conditions representative of those presented can be expected at the PORV inlets as well.

3-71

listed in Table 3-32, is 2535 psia {reference 3, Table 5-2). From the same table, the range of expected liquid temperatures for these units is defined as 613.4 to 632. 7°F.

Two (2) water seal simulation and two (2) steam tests were perfonned on the Garrett relief valve with inlet opening pressures (2755, 2760, 2760, and 2760 psia) 1n excess of the maxi111.1111 predicted (2752 psia) for FSAR events (see Table 3-31 9 tests 105, 106, 97, and 98}o

To address transition and 1;quid conditions resulttng from FSAR events in these units one test of each type was perfonned. The transition test (see Table 3-31, test 104) was perfonned at a pressure of 2760 psia with 653°F water following the steam initially at the valve inletu The liquid test (see Table 3-31, test 103)

was performed with an opening pressure of 2758 psia and a temperature of 648°Fu For both tests, the liquid tempeW"atures tested {648 and 653°F) were representative of these expected (613u7 to 632s7~F) and opening pressures (2758 and 2760 ps1a)

were greater than the maximum value expected (2535 psi a).

PORVs in Combustion Engineering and Westinghouse units have opening set points of 2400 and 2350 psia, respectively during power operation, (reference 2, Section 5o2o2 and reference 3, Table 3-2)o Asst.ming a minimum blowdown setting of 20 psi, the maximun closure pressure for relief valves in these units would be 2380 psia~ During the tests discussed above, the signal to close the Garrett Felief valve was given at inlet pressures ranging from 2210 to 2480 psia which exceed or are within 7o2 percent of the maxiirum iflooplant value (2380 psia) o

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Garrett relief valve was tested are considered adequate to reprea sent those expected for FSAR events in the PWR units listed in Table 3m32o


The maximum quasi-steady bending moment induced while the Garrett relief valve was in the opening or closing precess during a •preloada test was 34~000 inmlb (see Table 3m3l, test 98)~


As discussed fn Section 2e2o4, it fs important to demonstrate the Garrett relief

"" valve's ability to operate under developed back pressures at least as high as

•

those expected in ·service. Full pressure steam tests were perfonned during Marshall testing of this valve with back pressures as high as 825 psia (see Table 3-30, tests 5 and 6). In addition, all liquid, transition, and water seal simulation tests ( see Table 3-31, tests 99, 100, 101, 102, 103, 104, 105, and 106) were perfonned with a discharge pipe orifice which had been demonstrated to develop back pressures of between 580 and 623 psia under full pressure steam flow conditions {see Table 3-31, tests 97 and 98). Therefore, the back pressures developed during liquid, transition and water seal simulation testing of the Garrett relief valve correspond to those expected under similar flow conditions in PWR units with predicted steam back pressures in the range of 580 to 623 psia.



Extended High Pressure Injection events result in steam followed by water discharge through the relief valves in the Westinghouse units {reference 3, Section 5-3) and no relief valve discharge in the CE unit (reference 2, Sec-tion 5.2.2) listed in Table 3-32. Liquid discharges at the PORV opening set point. in these units (2350 psia) with temperatures in the range of 565~to 569°F are predicted (reference 3, Table 5-3).

To represent the conditions· resulting from such events, one transition and two (2) liquid tests were perfonned. The liquid tests (see Table 3-31, tests 99 and 103) were perfonned at opening pressures exceeding 2350 psi a and temperatures ranging from 438 to 648°F. As discussed in Section 3.11.1, a transition test (see Table 3-31, test 104) was also perfonned with an opening pressure above 2350 psia and a liquid temperature of 653°F. Closure pressures for all tests discussed above exceeded that expected in the Westinghouse units listed in Table 3-32 (2330 psia) with the exception of test 99, a 438°F water test. The closure pressure for this test {2030 psia) was 13 percent below maximum expected closure pressure.

This valve has a relatively high flow capability and depressurized the test system rapidly. The minimum test duration required to obtain quasi-steady fl ow data was approximately four (4) seconds. Actuation of this valve under -450°F water conditions at the maximum allowable test pressure (2760 psig) and allowing a four (4) second discharge, resulted in closure pressures somewhat below that/expected in service •

3-73

Although thi.s test closure pressure was 1 ower than desired, the other 1 i quid test (test 103) had a closure pressure (2480 psia) well above the target value and is considered adequate to demonstrate closure capability under the expected liquid conditions.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Gan-ett relief valve was tested are considered adequate to represent those expected for Extended High Pressure Injection events in the units 1 i sted in Table 3-32.


See Section 3.llol.


See Section 3oll.lo


The range of expected fluid conditions for Cold Overpressurization events was provided via the Westinghouse Plant Condidons Justification Report (reference 3) only for those Westinghouse units as noted in Table 3-32, ioe., only for the Wolf Creek ·and Callaway l and 2 unitso 'Therefore, justification that such conditions were addressed in the testing of the Garrett relief valve is provided only for these units and the Sto Lucie 2 unit which was addressed in reference 2 (see Section 1 .. 3 for discussion) e Such justification for the other units 1 isted wil 1 be included as part of theiF plant0 specific eva1uationso

INLET FLUID STATE/PRESSURE/TEM>ERATURE

The full range of potential Cold Overpressurization conditions for all Westinghouse units evaluated is depicted in Figure 5ml of reference 3e As discussed in Section 5o4 of reference 3, steam9 transition» and subcooled water conditions occurring over a large range of pressure and temperatuFe are possible during such events in the Westinghouse units evaluatedo

From Figure 5-1, reference 39 it is noted that full pressure (2350 psia) liquid discharges are expected in Westinghouse units over a temperature range of 250°F to 650°F. Based on this figure and discussions in Section 5.4 of reference 3, it is noted that reduced pressure (300 psig) liquid discharges are expected in

3-74 •

Westinghouse units over a temperature range of 100 to 400°F. In addition, it is noted from Table 5-27, reference 2, that reduced pressure {477 psia) liquid discharges are expected in St. Lucie 2 over a temperature range of 100 to 417°F.

The full pressure (greater than 2350 psi a) steam and transition tests previously discussed (Table 3-31, tests 97, 98, and 104) envelop expected steam and transition conditions resulting from such events. ·

Liquid tests over a range of pressures and temperatures representative of those shown in Figure 5-1, reference 3, were also performed. Full pressure (greater than 2350 psia) tests with liquid temperatures ranging from 249 to 648°F were performed (see Table 3-31, tests 102 and 103). In addition, reduced pressure (-685 psia) tests with liquid temperatures ranging from 104 to 447°F were performed (see Table 3-31, tests 100 and 101).

Valve closure pressures for one of the full pressure liquid tests discussed above (test 103) exceeded the maximum expected closure pressure (2330 psia). However, the closure pressure for test 102, a 249°F water test was 1880 psia which is 19.3 percent below the maximum expected value for Westinghouse units. As discussed in .Section 3.11. 2, this test condition represented the maximum capability of the test facility. Valve closure pressures for all reduced pressure tests were well above the minimum expected opening pressure (300 psia) corresponding to the lower end of the expected liquid temperature range (100-400°F) in Westinghouse units and were above the peak predicted pressure {477 psia) in St. Lucie 2~

~though, due to test facility constraints, the closure pressure on one of the full pressure liquid tests was somewhat below the in-plant value, in total, the tests performed to represent this event (tests 100, 101, 102, and 103) resulted in inlet fluid conditions (state, pressure, and temperature) which are considered adequate to represent those expected for Cold Overpressurization events in the St. Lucie 2, Wolf Creek, and Callaway 1 and 2 units.


See Section 3.11.1.


See Section 3.11.1.

3-75

w I '..! 0\

•

ltl'!U/ IWlSHAU. PORV UST DATA (J)

TMU l - JO "AS nsnom MRSl!AU. nu MURI 11'1111 ~It( 6ARRETI If UllT WAUIE

~OIDIVSON$ av WALVE QPf!ftMS 1tAu1 mn

JEST '!EST RUll!l il'RESS. 1( ... 110. iWPI (PSll\) (Or)

SifM SUM r.u5 ~SAT.)

1 STUM SJEMI 205 (SAT.) J STEM STEAlt HIS (SAT.)

4 SiEM SHAH 2«15 (SAT.)

i SUM :nw 2425 (SAi.)

' SYEM STENO Z«SO (SAi.)

1 SUNf STfAll 1465 (SAT.)

I SifM SUM 2455• · (SAll'.i 9 SUM SUMI 24«5 (SAV.»

10 Si(All SUAlll 24l!!i SSAi.B

Bl SUM 5ifAl!I HU «SAT.)

IOUS: (I) l!liai111111 Qu1&9 sae11d1 ~Osch~rge pigie prreSiure. (2) Plot recorded.

HI ACCilllllATOA DUllU111 .. FLUID i'RESS. ltHP. gsu:) (PSIA) (Or)

~ A$ VALVE INlfT 11

2ll

H

H

JC

H

15

H

96

if

(2)

TIANSlfll CONDITBOIS

mn. 1aM1 r.nr=Jo SIGNAL 51V£1 PIPE PUSS. JO CLOSE VALllE (PSHA)

PSIA

19!15 1115

llO:Hi 1120

1015 115

20]0 IH 2050 1125

2015 1125

2075 JJ5

2055 355

2055 :MO

2055 JU

2015 1115

h) Ve1i rrHuh& 11rre for 111v11'iuetiolll aeus 011ly. Vote~ of 66 iluppie•niary 11sl11e 1ctu1tton cyc1e11 wni gNtrfor11ed under sl•H•ir condtUont •

(E .E PORV TEST DATA)

TABLE 3 - 31 "AS TESTED" WYLE. PHASE Ill TEST MATRIX FOR THE

GARRETT RELIEF VALVE

INITIAL CONDIJIONS TRANSIENT CONDITIONS

Valve Inlet

At Valve Inlet In Accumulator Pressure Haxl111um Maximum Maximum ------ at Discharge Bending Valve Test Closure Pipe Moment Acceleration

Temp Press. Temp Press. Duration Signal Press.llJ lnduced(3)1nduced J~No. Test Type fluid l~U 11!!!.!l Fluid (Ofl_ 1esla) _(seconds) (es la) !1?.!!.!lill (tn:!!1_ .~g's)

97-GA-lS Steam Steam 603 2,760 Steam 682 2,760 4 2,346 580 N/A 8.1

90-GA-2S Steam Steam 603 2~760 Ste an\ (Pre load)

679 2,760 4 2,275 623 34,000 0.4

99-GA-311 Water Water 438 2,760 Water 461 2,760 4 2,030 485 N/A 1.2 w I

-...J l00-GA-4W Water Nater 447 683 Water 460 603 12 610 255 H/A l.2 -...J

IOI-GA-SW Water Water 104 606 Water 94 1586 11 495 25 N/A 4.2

l02-GA-6W Water Nater 249 2,640 Water 250 2,1540 3 l,880 92 N/A 1L8

103-GA-711 Water Water 648 • 2, 750 Water 653 2,758 3 2,480 780 N/A 4.0

104-GA-OS/ll Transition Steam 602 2,760 Water 653 2,760 6 2,420 800 N/A 5.2

105-GA-911/W Water Nater 293 2,755 Water 651 2,755 16 2,225 875 N/A 5.8 Seal Simulation

106-GA-lOW/W Water Water Seal

130 2,760 Water 650 2,160 16 2,210 860 H/A 8.7

Simulation

Notes: (1) Values shown were measured 50" downstrea1n of v.alve exit (same locatlon as for all other PORV's tested). For this valve an additional discharge pipe pressure sensor (PS-6) was mounted lmnedlately downstream or the valve, (see section 5.0 for plot of observed pressure).

(2) Ho back pressure orifice was used 1n the Garrett PORV testing.

(3) Value shown corresponds to maximum moment applied while valve was In the opening/closing process.

Table 3-32

PWR UNITS WITH GARRETT RELIEF VALVES

.Y!lll St Lucie 2

Alvin W. Vogtle 1~ 2 (GAE, GBE)l

Wolf Creek (SNUPPS)

Millstone 3 (NEU)l

Callaway 1, 2 (SNUPPS)

NSSS Vendor

Combustion Engineering

Westinghouse

Comments

Four 1 oop units

Four 1 oop unit

Four loop unit

Four 1 oop units

lJustification of conditions resulting from Cold Overpressu~ization events is not presented for these units. Such justification will be provided as part of these Utilities• plant-specific evaluations.

3-78

•

Section 4.0

JUSTIFICATION OF SAFETY VALVE TEST CONDITIONS

4.1 GENERAL

4.1.l DEFINITION OF A LIMITING SET OF SAFETY VALVE INLET FLUID CONDITIONS

As discussed in section 2.1, a limiting set of safety valve fluid inlet conditions enveloping those expected in all PWR units listed in Table 1-1 was developed and is presented in Table 4-1. These conditions were developed based on ongoing discussions with the three PWR NSSS vendors as well as review of the "Plant Condition Justification Reports" (references 1, 2, and 3) provided by each vendor. The following provides justification for the 1 imiting nature of the values presented in Table 4-1.

FSAR EVENTS

FSAR events result in steam discharge in all B&W and Combustion Engineering units (reference 1, Table 5-1 and reference 2, Tables 5-1 through 5-27) and steam, water, and transition conditions in some Westinghouse units listed in Table 1-1 (reference 3, Table 5-2). In addition, many of the units listed in Table 1-1 utilize a loop seal inlet piping configuration.

The maximum predicted inlet pressure for FSAR events resulting in steam discharge in these units is 2760 psi a (reference 2, Table ~~17) and the maximum pressurization rate for such events is 240 ps.i/s~c (l"eference 3, Table 5-1). . . .

The maximum pressure predicted during FSAR events resulting in transition or liquid discharge in these units is 2575 psia (reference 3, Table 5-2). The same table defines the range of predicted liquid temperatures and the maximum pressurization rate for such events in these units as 554° to 672°F and 12.2 psi/sec, respectively.

4-1

EXl~NDED HIGH PRESSURE INJECTION EVENTS

Extended High Pressure Injection events result in steam, transition, and liquid discharge with peak pressures at or slightly above the safety valve opening set point (2507-2515 psfa) in B&W (reference 1, Tables 5-2 and 5-3) and Westinghouse 4-loop (reference 3, Table 5-3) units. In addition, similar conditions are predicted for Maine Yankee (reference 2, section 5c2.1). Pressurizatian rates for such events can range fram 0 to as high as 65 psi/sec (reference 19 page 4e34). The range of liquid temperatures predicted for the units listed in Table lel is ~nveloped by that predicted (400 to 650°F) for B&W units (reference 1, Tables 5-2 and 5-3).

It should be noted that although Extended High Pressure Injection events can result in steam, water, or transition flow, tests selected to represent this event included only transition and water testse This was due to the fact that the steam conditions which might oceur during such events are enveloped by testing performed to address FSAR events.

COLD OVERPRESSURIZATION EVENTS

Cold overpressurization events do not challenge safety valves in the PWR units for which such analyses are discussed in references 19 2, and 3. Therefore, no discussion of the correlation between the conditions under which the selected safety valve designs were tested, and those resulting from such events, is presented. For those units not addressed by Cold Overpressurization analyses presented in f'eferences 1, 2, and 35 justification that the tested conditions represent those expected for such events in their units will be included in their plant specific evaluationso

OPENING/CLOSING PRESSURES

The safety valves tested were adjusted to open on steam at a set point of 2500 or 2515 psia using procedures consistent with those utilized in PWR unitso There.fore, the observed opening pressures of the valves tested are consistent with those expected in PWR units.

The closure pressure for the spring loaded safety valves tested is primarily a function of ring positions. Therefore, the observed closure pressure of the spring loaded safety valves tested are consistent with those expected for PWR valves of the

•

•

same design and with the same ring positions. ihe closure pressure for the pilot • ~--.--o-perated-safety-va~ ve-design-tested-(-T-arget--Rock-69G-)-is-not-adju-stab-1-e-a.nd-the--. - --- __

observed closure pressure is consistent with that expected in service.

4-2

INLET PIPING CONFIGURATIONS

Table 4-2 presents geometric details of the safety valve inlet piping configurations tested. Note that the test configurations have been assigned letter designations which correspond to those shown in the test matrices appearing in the following sections. As discussed in section 1.3, the inlet piping configurations tested were selected to be representative of in-plant installations. Justification of the applicability of the tested inlet piping geometries to specific inplant installations will be included as part of each PWR utility's plant specific evaluation.

4.1. 2 JUSTIFICATION OF TESTS PERFORMED WITH "REFERENCE" RING POSITIONS

As discussed in section 2.3.7, a complete matrix of tests, covering the full range of expected fluid conditions, was initiated for each spring loaded safety valve only after "reference" ring positions had been established. The detennination of "reference" ring positions was based on.initial tests perfonned under full pressure steam conditions. The following sections present the specific conditions under which each selected safety valve design was tested utilizing the "reference". ring positions* and justification that the fluid inlet conditions for these tests adequately represent those expected during FSAR and Extended High Pressure Liquid Injection events in the PWR units_listed in Table 1-1. Justification of the

•

applicability of these tests to in-plant valve installations (including the specific in-plant ring positions utilized} will be provided as part of each utility's plant-specific evaluation. Other conditions tested, including the effects of discharge piping (back pressure and bending moment) are also discussed.

*As the pilot operated valve design tested (Target Rock 69C} has no ring adjust-• ment, the complete matrix of tests perfonned on this valve 1s presented.

I

4-3

~ 8 ~

•

FSAR EVENTS

0 RESULTING Hf STEAM DISCHARGE

0 RESULTING KN LIQUID DISCHARGE

EXTENDED HIGH PRESSURE INJECTION EVENTS

COLD(l) OVERPRESSURIZATION EVENTS

Jable 4-1

LIMITING RANGE OF SAFETY VAlYE INLET flUIO CONOiTIONS FOR ?WR UNITS liSlEO IN TABLE 1-1

MAXIMUM RANGE OF RANGE OF SAFETY VALVE PRESSURIZER PRESSURIZATION

OPENING PRESSURES PRESSURE RATES PSiA PSIA ?SI/SEC

{2500-2515)i:11 2760 0-240

(2500-2515}±li 2575(2) O-l2a2

(2500-2515 )ill 2515 0-65

N/A NIA N/A

POSSIBLE RANGE OF LIQUID FLUID TEtt>ERATURES AT

STATES ON VALVE INLET OPENING Of

STEAM OR N/A LOOP SEAL

STEAM OR 554-672 LOOP SEAL OR TRANS IT ION OR LIQUID

~,

STEAM OR 400-650 LOOP SEAL OR TRANSITION OR LIQUID

N/A N/A

(1) Cold Overpr2ssurization events do not challenge safety valves in the PWR units evaluated for such events 1~ references 1

9 2

0 and 3e 1herefore0 no specific safety valve testing was perfonned to cover

poss1b1e conditions resu1t1ng from such events.

(2) Artif~c1a11y high due to assuned open1ng at 2575 ps1a. Expected in-plant value would be at actual opening pressure (nominally 2500-2515 ps1a)o

• ••

•

CONFIGURATION DESIGNATION

"A" LENGTH

I.D.

"B" LENGTH

"C" LENGTH I.D.

"O" LENGTH I.D.

"E" LENGTH I.O.

"F" LENGTH I .D.

"G" LENGTH laD.

"H" LENGTH I.D.

"I" LENGTH I .O.

VESSEL

Table 4-2

SUMMARY OF SAFETY VALVE INLET PIPING CONFIGURATIONS TESTED

INLET PIPE COMPONENTS

NOZZLE VENTURI REDUCER PIPE REDUCER PIPE REDUCER FLANGE - -17 38 12 115.6 N/A N/A N/A N/A

~.813 6.813 6.813/ 4.897 4.897

17 38 6.813 6.813

17 38 6.813 . 6.813

17 6.813

17 6.813

17 6.813

17 6.813

17 6.813

17 6.813

38 6.813

38 6.813

38 6.813

38 6.813

N/A N/A

38 6.813

N/A N/A

N/A N/A

N/A N/A

N/A N/A

N/A N/A

N/A N/A

N/A N/A

N/A N/A

6 6 N/A N/A 6.813 6.813/

40897

11 10 N/A N/A 6.813 6.813/ N/A N/A

2.125

11 6 6.813 . 6.813/

3.152

6 10 6.813 6.813/

2.624

6 6 6~813 6.813/

3.152

13 6 6.813 60813/

4.897

9 6 60813 6.813/

5.189

33 6 6.813 6.813/

4.897

97.7 4 3.152 3.152/

2.125

N/A N/A N/A N/A

91.7 4 3.152 3.152/

2.624

104.5 N/A 4.897

76 N/A 5.189 N/A

116. 5 N/A 4.897 N/A

11

6 2.125

6 2.12s

11 2.624

7 2.624

10 40897

7 5.189

7 4.897

NOTE: 1

All values shown are in inches • I I

i

4-5

- -····· ... - ... . _ .... -··· .. ...!.. .. ··-·-- ._..._ .... _ .. ______ .. !-.~._..;..~ .. ~ ...... =---····- .. _____ .,. ___ .. _ ..... --.... --- .. ·-··· ,, .. .. . .

•

•

••

4.2 DRESSER 31709NA

Table 4-3 presents the conditions under which the Dresser 31709NA safety valve was tested utilizing "reference". ring positions. The following sections provide specific justification that for each event type considered (FSAR and Extended High Pressure Injection), these conditions are representative of the limiting conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability (back pressure and bending moment), imposed on this valve design, are discussed.

4.2.1 FSAR EVENTS

INLET FLUIO CONDITIONS STATE/PRESSURE/TE"1PERATURE

Testing was perfonned on the Dresser 31709NA safety valve with both a short and a long (loop seal) inlet piping configuration (see Table 4-2 for configuration details).

Short Inlet Testing

Two {2) steam tests {see Table 4-3, tests 615 and 620) were perfonned on the Dresser 31709NA vaJve with "reference" ring positions at peak p~essures (2639 and 2667 psia) and pressurization rates (317 psi/sec) representative of the maximum predicted values (2760 psia and 240 psi/sec), for FSAR events in the PWR units listed in Table 1-1.

To address the transition and liquid conditions resulting from FSAR events, one {l) transition and two (2) liquid tests were perfonned. These tests {see Table 4-3, test 628, 630, and 1308) were perfonned at peak pressures (2530, 2393 and 2513 psia) and pressurization rates (2.7, 2.5 and 1.8 psi/sec) representative of those expected (2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test (""656°F) was representative of the upper end of the expected range (554 to 672°). The liquid temperatures (587 and 535°F) for the liquid tests represent the low end of the predicted li~uid temperature range.

Long Inlet Testing

Only one {l) test (see Table 4-3, test 201) was perfonned on the Dresser 31709NA valve with a long inlet piping configuration. No additional long inlet testing was perfonned since the only PWR units with this valve design and a long inlet elected to shorten their inlet piping as .a means of improving valve·perfonnance.

i

4-6

... :-~ .. ·- ···--· .... ·· .. ---··-··- .......... __ ....... ~ ...... - ~ .. - , .. . ...... ···-·"'··--

Based on the above, sufficient testing was perfonned to assess the operability·of the Dresser 31709NA safety valve over a range of inlet fluid conditions representative of those expected for FSAR events in the PWR units listed in Table 1-1.


As discussed in section 2e3o4 1 it is considered important to demonstrate safety valve operation at back pressure' at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table l·l) were performed on the Dresser 31709NA with •reference" ring positions with valve outlet pressures as high as 345 psia (test 1305).

BE~DING MOMENT INDUCED

The maxi1111m quasi-steady bending moment imposed on the Dresser 31709NA safety valve while the valve was in the opening or closing process was 473,200 in-lb (see Table 4-3, test 1308).

4o2e2 EXTENDED HIGH PRESSURE INJECTION EVENTS

INLET FLUID CONDITIONS STATE/PRESSURE/TEMPERATURE

Short Inlet Testing

To address possible transition and liquid ~ondit1ons resulting from Extended High Pressure Injection events, one (1) transit;on and three·{J) liquid tests (see Table 4q3, tests 628, 630, 1308, and 1311) were performed at peak pressures (2530 9

2393 2 2513. and 2558 psia) and pressurization .-ates (2.7 9 2.,5 9 1.8, and 2GG psi/~ec) representative of those p~edicted (2515 psia and 0-65 psi/sec) for sueh events. The range of liquid temperatures under which these tests were performed (429 to -656°F) is rep.-esentat;ve of the full range of predicted temperatures (400 to 650°F).

Long Inlet Testing

See section 4.2.lo

Based on the above, sufficient testing was performed t~ assess the operability of the Dresser 31709NA safety valve over a range of inlet fluid eandit1ons represen:. tative of those expected for Extended HighmPressuFe Injection events in the PWR units listed 1n Table 1-1.

•

•

--------------------- -------------------- --- ---·---

4-7

•

•


During transition and liquid testing of the Dresser 31709NA safety valve, with a short inlet, the same back pressure orifice was used in the discharge piping as was used during full pressure steam tests in which back pressures from 326 to 345 psia were attainedo Tilerefore, the back pressures developed for these liquid tests correspond to those expected under similar flow conditions in PWR units with expected steam back pressures in the 326 to 345 psia range.


See section 4.2.1 •

4-8

.,.. 9

'°

I

TEST TEST

Table 4-3

00AS TESTED~ COMBUSTION ENGUNEERING MATRHX Of TESTS PERFORMED OH1JHE DRESSER 31709NA SAFETY VAUE WITH 11REFERENCE19 RING POSITIONS{

INLET l?llPING CONDITIONS JUST PRIOR TO NO. TYPE CONFIGURATION(2) VALVE OPENING TRANSIENT CONDITIONS

PEAf AT VALVE iNLEl IN PEAK PEAK BENDING

PRESS ACCUKJLATOR TANK(l) BACK MOMENT FLUID TEW PRESS RATE TEN» PRESS PRESSURE INDUCEo(4)

Of PSIA PSI/SEC FLUID~ PSIA PSIA IN-LB

201 STEAM °'A" STEAM n 2486 400 STEAM n 2680 (5) 1370500

615 STEAM ilen STEAM 1 2568 317 STEAM 1 2639 326 100.100 620 STEAM STEAM 7 2540 317 STEAM 1 2667 194 95,550 628 TRANSlllON STEAM 1 2530 2o7 STEAM/

~~:~:J WATER (7) 2530 910000 630 WATER WATER 589 2393 2o5 WATER 625 2393 100,100

1305 S.TEAH STEAM fl) 2530 308 STEAM (7) 2652 345f 6J

200,200 1308 WATER WATER 535 2487 le8 WATER 562 2513 145 6 413,200 1311 WATER WATER 429 2558 2.6 WATER 415 2558 100 445.900

.(1). The ~tng pcstt1ons for these tests were: upper (-48) 1 middle (-20) 9 lower (0) for all tests except test 201e Test 201 positions were: upper (-48), middle (+34)~ lower (-20). and were not considered Mreference• positions.

(2) See Table 4-2 for configuration details. (3) Valve inlet pressures slightly lower due to frictional pressure drop tn inlet p1p1ng. (4) The maximum bending moment imposed during a11 31709NA tests was 473.200 1n-lb (test 1308). (65) Measurement unreadable due to valve chattere

( ) Same 8P orifice used as for test 6150 (7) Test initiated with saturated steam or water nominally at 2300 psta 11 1 ae., at a temperature of -656°F.

I

i ,. I

~·

1

•

4.3 DRESSER 31739A

Table 4-4 presents the conditions under which the Dresser 31739A safety valve was tested utilizing 11 reference 11 ring positions. The following sections provide specific justification that for each event type considered (FSAR and Extended High Pressure Injection), these conditions are representative of the limiting condi.tions presented in Table 4-1. In addition, the discharge piping effects on valve operability (back pressure and bending moment), imposed on this valve design are discussed.

4.3.1 FSAR EVENTS

INLET FLUID CONDITIONS STATE/PRESSURE/TEr-PERATURE

Testing was performed on the Dresser 31739A safety valve with both a short and a long (loop seal) inlet piping configuration (see Table 4-2 for configuration details).

Short Inlet Testing

Three (3) steam tests (see Table 4-4, tests 316, 318, and 1104a) were performed on the Dresser 31739A valve with 11 reference11 ring positions at peak pressures (2703, 2685, and 2720 psia) and pressurization rates {320, 285, and 316 psi/sec), representative of the maximum predicted values {2760 psia and 240 psi/sec) for FSAR events in the PWR units listed in Table 1-1.

To address the transition and liquid conditions resulting from FSAR events, one (1) transition and two {2) liquid tests were performed. These tests (see Table 4-4, test 1107, 1110, and 1112) were performed at peak pressures (2489, 2521, and 2393 psia) and pressurization rates (2.8, 2.3 and 3.1 psi/sec) representative of those expected (2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test (-656°F) was representative of the upper end of the expected range (554 to 672°F). The liquid temperatures (570 and 493°F) for the liquid tests represented the lower end of the range of predicted liquid temperatures.

Long Inlet Testing

One (1) steam and one (1) loop seal test were performed on the Dresser 31739A valve with 11 reference" ring positi6ns and a long inlet confi~uration (see Table 4-4, tests 1018 and 1021). pie peak pressures {2657 and 2698 psia) and . pressurization rates (308 and 329 psi/sec) were representative of those predicted (2760 psia and 240 psi/sec) for FSAR events in the PWR units listed in Table 1-1.

4-10

To address possible transition and liquid conditions resulting from FSAR events, one (1) transition and two (2) liquid tests were performed (see Table 4-4, tests 1025, 1027, and 1030). These tests were performed with peak pressures (2536, 2363, and 2458) and pressurization rates (2.0, 2.4, and 1.9 psi/sec), representative of those expected (2575 psia and O-l2o2 psi/sec) for such events. The range of liquid temperatures covered in these tests (515 to -656°F) is representative of the predicted temperature range (554 to 672°F)o

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Dresser 31739A safety valve was tested utilizing "reference" ring positions, on both a short and a long inlet pipe, are considered _adequate to represent those predicted for FSAR events fn all PWR units listed in Table 1-1.


As discussed f n section 2.,3.4lil ft is considered important ta demonstrate safety

valve operation at back pressures at least as high as those expected 1n service~ Full pressure steam tests (representative of expected FSAR events for those units 1 isted in Table 1-1) were performed on the Dresser 31739A with "reference" ring positions with valve outlet back pressures as high as 866 psia (short inlet) and 586 psia (long inlet).


The maximum quasi-steady bending moment imposed on the Dresser 31739A valve while it was fn the opening or closing process was 244 9650 ine1b (see reference 4, Table 301.lab, test 1011).,

4Q3a2 EXTENDED HIGH PRESSURE INJECTION EVENTS

INLET CONDITIONS STATE/PRESSURE/TEM>ERATURE

Short Inlet Testing

To address possible transition and liquid conditions resulting from Extended High Pressure Injection events 9 one (1) transition and three (3) liquid tests were peFfonned. These tests (see Table 4-4p tests 1107~ 1110~ lll2D and 1114) were perfo~ed at peak pressures (2489, 2521, 2393 9 and 2749 psia) and pressurization rates (2.8, 2.3, 3ol, and 3.2 psi/sec), representative of those predicted (2515 psia and 0-65 psi/sec) for such events. The range of liquid temperatures under which these tests were performed (407 to -656°F) is representative of the full range of predicted temperatures (400 to 650°F)o

4-11

Long Inlet Testing

To address possible transition and liquid conditions resulting from Extended High •

•

Pressure Injection events, one (1) transition and two (2) liquid tests were performed with a long inlet piping configuration, (see Table 4-4, tests 1025, 1027 and 1030). The transition test was perfonned with a water-filled loop seal. The

•

peak pressures (2536, 2363, and 2458 psia) and pressurization rates (2.0, 2.4 and 1.9 psi/sec), are representative of those predicted (2515 psia and 0-65 psi/sec) for such events. The range of temperatures under which these tests were performed (515 to ...056°F), is representative of the upper end of the predicted temperature range (400 to 650°F) for such events. No testing of this valve, on a long inlet pipe, at temperatures lower than 515°F was perfonned because of the expectation that the valve's performance would be similar to that observed under 515°F water flow conditions.

Based on the above, the range of fluid inlet conditions under which the Dresser 31739A valve was tested utilizing "reference" ring positions, on a short inlet pipe, adequately represents those expected during Extended High Pressure Injection events in all PWR units listed in Table 1-1. In addition, sufficient long inlet testing was performed on this valve to assess its operability over the full range of expected fluid inlet conditions in these units.

RACK PRESSURE DEVELOPED

During transition and liquid testing of the Dresser 31739A safety valve, with a short and a long inlet, the same back pressure orifice was used in the discharge piping as was used during full pressure steam tests in which back pressures from 570 to 600 psia were attained. Therefore, the back pressures developed for these liquid tests, correspond to those expected under similar flow conditions in PWR units with expected steam back pressures in the 570 to 600 psia range.



4-12

TEST NO.

316 318 320

~ 322 e 1018 _. w 1021

1025

1027 1030

(l) (2) (3) (4)

!5) 6)

•

TEST TYPE

Table 4-4

GIAS TESTE0'° COMBUSTION ENG!NEEIHNG MATRIX OF TESTS PERFORMED OJt )THE DRESSER 31739A SAFIETY VALVE wrrn °1REFERENCE11 RING POSITIONsll

INL!ET PIPING CONDITIONS JUST PRIOR TO CONfiGURATION(2) VALVE OPENING TRANSIENT CONDITIONS

PEAK AT VALVE INLET IN PEAK PEAK BENDING

BACK MOMENT PRESS ACCUrt.llATOR TANK(3) FLUID lErt> PRESS RATE TEW PRESS PRESSURE INOUCED(4)

Of PS IA PS I /SEC FLU l 0 _1_ PSIA PSIA IN-LB

STEAM 11cee STEAM r 2590 320 STEAM 6 2703 195 89,538

STEAM

' STEAM 6! 2483 285 STEAM 6 2685 195 94,250

STEAM STEAM 6 2580 316 STEAM 6 2667 866 103,675 STEAM + STEAM 6 2530 311 STEAM 6 2670 609 103,675 STEAM 0101111 STEAM (6 2455 308 STEAM 6 2657 570 87,375 LOOP SEAL WATER 104 2582 329 STEAM 6 2698 586(5) 157,275 LS-TRANSITION WATER 104 2525 2.0 STE AW 6) 2536 781 15711275

WATER 6) 590(5) WATER WATER 618 2350 3.2 WATER 621 2363 133,975

WATER WATER 515 2408 1.8 WATER 522 2458 64o(5) 87 11375

The r1ng positions for these tests were: upper (-48), mtdd1e (-40) 0 lower (+11). See Figure 4-2 for configuration details. Valve inlet pressures slightly lower due to frictional pressure drop 1n inlet p1p1ng. The max1mtJD bending moment imposed on the Dresser 31739A during the opening or closing process was 244 0 650 in-lb (see reference 40 Table 3.1.1.~ 0 test 1011). . Same !back pressure oidfice used as for tests 1018 and 1021. Test 1ntttated with saturated steam or water nominally at 2300 ps1a, 1.e., at a temperature of ~656°F •

••

·'

•

~ • _, .i:o.

TEST TEST NO. TYPE

• Table 4-4 (con't)

11AS TESTED 11 COMBUSTION ENGINEERING MATRIX OF TESTS PERFORMED OR)THE DRESSER 31739A SAFETY VALVE WITH "REFERENCE" RING POSITIONSl 1

INLET PIPING CONDITIONS JUST PRIOR TO CONFIGURATION(2) VALVE OPENING TRANSIENT CONDITIONS


BACK MOMENT

• ••

PRESS ACCUt1.ILATOR TANK(J) FLUID TEMP PRESS RATE UMP PRESS PRESSURE INDUCE0(4)

Of PSIA PSI/SEC FLUID-~ PSIA PSIA

1104a STEAM "C" STEAM (6) 2550 316 STEAM ~6J 2720 600 1104b STEAM STEAM ~6~ 2240 0.1 STEAM 6 2240 94 1107 TRANSITION STEAM 6 2489 2.8 STEAM/

725(5) WATER (6) 2489 1110 WATER WATER 570 2521 2.3 WATER 608 2521 500(5) 1112 WATER WATER 493 2387 3.1 WATER 513 2393 290rn~ 1114 WATER WATER 407 2470 3.2 WATER 414 2749 211

(1) The ring positions for these tests were: upper (-48), middle (-40), lower (+11). (2) See Table 4-2 for Inlet Configuration details. (3) Valve inlet pressure slightly lower due to frictional pressure drop in inlet piping.

IN-LB

230,913 (7)

226,200 169.650 158,340 84,825

------(4)-- The maximum bending moment imposed on the Dresser 31739A during the opening or closing process was 244,650 in-lb {see reference 4, Table 3.1.1.b, test 1011).

(5) Same back pressure orifice used as for test 1104a. · (6) Test initiated with saturated steam or water nominally at 2300 psia (i.e •• at a temperature of -656°F). (7) Not available.

4.4 CROSBY 3K6 (STEAM INTERNALS)

Table 4-5 presents the conditions under which the Crosby 3K6 (Steam Internals) safety valve was tested utilizing 11 reference11 ring positions. These tests were performed only with a short inlet piping configuration since the operability of such valves with a long inlet was demonstrated by tests of the Crosby 3K6 (Loop Seal Internals) safety valve on a long inlet (see section 4.5). The following sections -provide specific justification that for each event type considered (FSAR and Extended High Pressure Injection), the conditions under which t.his valve design was tested are representative of the limiting fluid inlet conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability (back pressure and bending moment), imposed on this valve design are discussed.

4.4.1 FSAR EVENTS

INLET FLUID CONDITIONS STATE/PRESSURE/TEr.PERATURE

Three (3j steam tests (Table 4-s,· tests 416, 425 and 441) were performed on the Crosby 3K6 {Steam Internals) safety valve with 11 reference11 ring positions at peak pressures.{2700, 2730 and 2700 psia) and pressurization rates (311, 325 and 289 psi/sec) representative of the maximum predicted values {2760 psia and 240 psi/sec) for FSAR events in tne PWR units listed in Table 1-1.

To address the transition and liquid conditions resulting from FSAR events, one (1) transition and three (3) liquid tests were performed. These tests (see Table 4-5, tests 428, 431a,* 435, and 438) were performed at peak pressures (2548, 2349, 2582, and 2490 psia) and pressurization rates (2.7, 1.8, 1.7, and 2.3 psi/sec), representative of those expected (2575 psia and 0-12.2 psi/ sec) for such events. The liquid temperatures for the liquid tests (622, 510, and 538°F) are representative of the full range of predicted liquid temperatures {554 to

.672°F).

Based on the above, sufficient testing was performed to assess the operability of the Crosby 3K6 (Steam Internals) safety valve over a range of fluid inlet conditions representative of those expected for FSAR events in all PWR units listed in Table 1-1.

*A subsequent actuation occurred during test 431 (43lb). As data for this test is incomplete, it is not discussed herein. . I

4-15


As discussed in section 2.3.4, it is considered important to demonstrate safety valve operation at back pressures at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table 1-1) were perfonned on the Crosby 3K6 (Steam Internals) valve, with •referencen ring positions, with valve outlet pressures as high as 705 psiao


The maxi111J111 quasi-steady bending moment imposed on the Crosby 3K6 valve {steam or loop seal internals) while it was in the opening or closing process was 161,500 in-lb (see Table 4-5, test 441).


INLET FLUID CONDITIONS STATE/PRESSURE/TEM>ERATURE

To address possible transition and 11quid conditions resulting from Extended High Pressure Injection events, one (1) transition and three (3) liquid tests weFe per"' fanned. These tests {see Table 4-5, tests 428, 431, 435, and 438) were perfonned at peak pressures (2548, 2349, 2582, and 2490 psia) and pressurization rates (2.7, 1.8, 1.7, and 2o3 psi/sec) representative of those expected {2515 psia and 0-65 psi/sec) for ~uch events. The range of liquid temperatures under which these tests were performed {510 to -656°F) 1s representative of the upper end of the maximt.an predicted range (400 to 650°F). No additional testing was perfonned on this valve design at lower liquid temperatures, due to the expectation that the valve's perfonnance would be similar to that observed during testing perfanned under SlD°F water flow conditions.

Based on the above, sufficient testing was performed to assess the operability of the Crosby 3K6 (Steam Internals) safety. valve over a range of fluid inlet ccmd'f..,

tions representative of those expected for Extended High Pressure Injection events in all PWR units listed in Table 1-1.,

BAO< PRESSURE DEVELOPED

During transition and liquid testing of the Crosby 3K6 {Steam Internals) safety va.lve, the same back pressure valve open percentage as was used during fu11 pres~

sure steam tests in which a back pressures ranging from 632 to 705 psia were attained. Therefore, the back pressures developed for these liquid tests corr~ spond to those expected under similar flow conditions in PWR units with expected steam back pressure in the 632 to 705 psia range.

4-16

•

•



•

•

4-17

laMe 4-5

0'AS TESTE011 COMBUSTION ENGJINEERING MATRIX Of llESTS PERFORMED ON THE CROSBY 3K6 (STEAM iNTERNALS) SAFETY VALVE WITH ~REFERENCE11 RING POSl!IONS(l)

TEST NO.

416 425 428

43la 43lb

. 435 438 441

TEST TYPE

STEAM STEAM TRANSITION

WATER WATER WATER WATER STEAM

!NlET PIPVNG . CONDKTIONS JUST PRIOR TO CONFIGURAll!ON(2) VALVE OPENING

~~~~~~~~~~~~~

AT VALVE INLET IN PRESS ACCUrt.ILATOR

FLUID TEM> PRESS RATE T[f'IS °F PSIA PSI/SEC FLUID °F

· STEAM (6) 2481 311 . STEAM (6) 2505 325

STEAM (6) 2548 2.7

WATER 622 2342 le8 WATER 616 2278 lo6 WATER 510 2454 1.7 WATER 538 2447 2a3 STEAM (6) 2473 289

STEAM (6) STEAM (6) STEAW WATER (6) WATER 631 WATER (7) WATER 520 WATER 554 STEAM (6)

(1) The ring posit~ons fo~ t~ese tests we~e: upper (-45), lower (-14)0

PEAK TANK(l) PRESS PSIA

2700 2730

2548 2349

f ~J 2490 2700

TRANSIENT CONDITIONS PEAK

PEAK BENDING BACK MOMENT PRESSURE INDUCEo(4)

PSIA IN-LB

705 140

:::~:J f ~~ 700 632

(7) 32,300

(7) 24,700 198000

1619500

(2) See Table 4-2 for configuration detatiso . (3) Valve inlet pressures s11ght1y.lowe~ due to frictional pressure drop in inlet piping. (4) Same BP valve setting (45% open) used for these tests as was used dull"1ng tests 416 and 44lo (5) The highest bending moment imposed on the Crosby 3K6 (steam or loop seal internals) valve while 1t was

1n the opening or closing process was 161 0500 in-lb. (67) Test 1n1t1ated with saturated steam or water nominally at 2300 psta (t.e.D at a temperature of -656°f).

( ) Not avatlablee · . (8) Data osci11atorye

•~

4.5 CROSBY 3K6 {LOOP SEAL'INTERNALS)

Table 4-6 presents the conditions under which the Crosby 3K6 {Loop Seal Internals}

•

.. safety valve was tested utilizing "reference" ring positions. These tests were perfonned only with a long inlet piping configuration since the operability of such valves with a short inlet was demonstrated by tests of the Crosby 3K6 (Steam Internals} safety valve on a short inlet {see section 4.4). The following sections provide specific justification that for each event type considered (FSAR and Extended High Pressure Injection), the conditions under which this valve design was tested are representative of the limiting fluid inlet conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability {back pressure and bending moment), imposed on this valve design are discussed.

4.5.1 FSAR EVENTS


One (1) loop seal test (Table 4-6, test 526) was perfonned on the Crosby 3K6 (Loop Seal Internals) safety valve with "reference" ring positions at peak pressure (2708 psi a) and pressurization rate (220 psi/sec), representative of the maximum predicted values {2760 psia and 240 psi/sec) for FSAR events in the PWR units 1 isted in Table 1-1. In addition, one (1) steam test {Table 4-6, test 535) was perfonned on this valve under conditions representative of those expected during a subsequent in-plant actuation after the loop seal has been cleared, i.e., with a reduced peak pressure and pressurization rate.

To address the transition and liquid conditions resulting from FSAR events, one (l} transition* test was perfonned. This test (see Table 4-6, test 532} was perfonned at a peak pressure {2573 psia} and pressurization rate {3.3 psi/sec) representative of those expected {2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test {-656°F) was representative of the upper end of the expected range {554 to 672°F). No additional liquid testing was perfonned on this valve design at lower liquid temperatures due to the expectation that the valve's perfonnance would be similar to that observed during testing performed under -656°F liquid flow conditions.

*The transition test was perfonned with a filled loop seal.

4-19

Based on the above, sufficient testing was performed to assess the operability of the Crosby 3K6 (loop Seal Internals) safety valve, over a range of fluid inlet conditions, representative of those expected for FSAR events in all PWR units listed in Table 1-1.

BAO< PRESSURE DEVELOPED

As discussed in section 2a394, it is considered important to demonstrate safety valve operation at back pressures at least as high as those expected in servicee Full pressure steam tests (representative of expected FSAR events for those units listed in Table 1-1) were performed on the Crosby 3K6 (loop Seal Internals) valve, with "reference" ring positions, with valve outlet pressures as high as 541 psia.


The maximum quasiasteady bending moment imposed on the Crosby 3K6 (Steam Loop Seal Internals) valve while it was fn the opening or closing process was 161,500 inalb (see Table 4-5, test 441).

4o5.2 EXTENDED HIGH PRESSURE INJECTION EVENTS


To address the transition and liquid conditions resulting from Extended High Pressure injection events 9 one (1) transition*, test was performed. 'This test (see Table 4-5, test 532} was performed at a peak pressure (2573 psia) and pressuri 0

zation rate (3o3 psi/sec) representative of those expected (2515 psia and 0~65 psi/sec) for such events. The liquid temperature for the transition test (..056°F) 9 was representative of the upper end of the predicted range of liquid temperatures (400m650°F). No additional liquid testing was performed on this valve due to the expectation that the valve's performance would be similar to that observed during testing performed under -656°F liquid flow conditions.

Based on the above, sufficient testing was perfoi"Tlled to assess the operability of the Crosby 3K6 (Loop Seal Internals) safety valve over a range of fluid inlet conditions representative of those expected for Extended High Pressure Injection events in all PWR. units listed in Table lale


4-20


During transition testing of the Crosby 3K6 {Loop Seal Internals) safety valve, the same back pressure orifice was used in the discharge piping as was used during full pressure steam tests in which back pressures from 471 to 515 psia were obtained. Therefore, the back pressure developed for this test, corresponds to that expected under similar flow conditions in PWR units with expected steam back pressures in the 471 to 541 psia range.


See section 4.5.1.

4-21

~ I IN N

TEST · TES1f

Table 4-6

"AS TESTEO":COMDUSHON ENGINEERING MATRIX Of TESTS PERFORMED ON THE CROSBY 3K6 (LOOP SEAL iNTERNAlS) SAFETY VALVE WKTH 11REfERENCEN RING POSiTIONs(l)

INLET PIPING · CONDITIONS JUST PRIOR TO NO. TYPE CONFIGURATION(2) VALVE OPENING TRANSIENT CONDITIONS


PRESS ACCUMJLATOR TANK(J) BACK MOMENT FLUID TEK> PRESS RATE TEMP PRESS PRESSURE(4) INDUCED(5) ~ PSIA PSI/SEC FLUID _1._ PSIA PSIA IN-LB

525 LOOP SEAL lllfl! WATER 110 2536 3.4 STEM n 2258 471 1011800

526 LOOP SEAL WATER 94 2608 220 STEAM 6 2708 513 147.500 529 lOOP SEAL WATER 86 2602 13.~ STEAM 6 2638 480 64.900 532 LS+TRANSITION WATER 360 2512 3.3 STEAM/ 6 I

WATER 2573 615 619950 535 STEAM STEAM (6) 2530 85.7 STEAM ~6i 2650 541 57.230 536 LOOP SEAL WATER 98 2637 43.6 STEAM 6 2677 507 64.900

(1) The ring positions for these tests were: upper (-115. relative to ~ottom of disc rtng) 0 lower (-14). {2) See Table 4-2 for conf1guraiton details. _ (3) Valve tn1et pressu~e slightly lower due to frictional pressure drop 1n inlet piping. (4) Same BP orifice used for all tests showne (5) The highest bending moment imposed on the Crosby 3K6 (steam or loop seal internals) valve while the

valve was in the opening or closing process was 161 8500 in-lb (see Table 4-5, test 441.) (6) Test tnittated with saturated steam or water nominally at 2300 psta (1.eo 9 at a temperature of ~656°f).

4.6 CROSBY 6M6 (STEAM INTERNALS)

Material properties of the internals for the PWR safety valves selected for testing are considered important from a valve seat leakage viewpoint but are not considered relevant to operability. Therefore, the operability of the Crosby a-16 valve with steam internals was adequately assessed during the testing of the Crosby 6'16 valve with loop seal internals (see section 4.7). Any internals wear which might be expected to occur in the Crosby 6M6 with steam internals valve was adequately assessed during testing of the Crosby 3K6 and 6N8 with steam internals (see sections 4.4 and 4.8).

4-23

•

4.7 CROSBY 6M6 (LOOP SEAL INTERNALS)

Table 4-7 pre~ents the conditions under which the Crosby 6M6 (Loop Seal Internals) safety valve was tested utilizing 11 reference11 ring positions. These tests were perfonned only with a long inlet piping configuration since the operability of such valves with a shorter inlet is expected to be as good or better than that observed with the test configuration. The following sections provide specific justification that for each event type considered (FSAR and Extended High Pressure Injection), the conditions under which this valve design was tested are represen-

. tative of the 1 imiting fluid inlet conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on this valve design are discussed.

4.7.1 FSAR EVENTS


One (1) steam and four (4) loop seal tests (Table 4-7, tests 1411, 929, 1406, 1415, and 1419} were performed on the Crosby 6M6 (Loop Seal Internals) safety valve with 11 reference11 ring positions at peak pressures (2664, 2726, 2703, 2760, and 2675 psia) and pressurization rates (300, 319, 325, 360, and 360 psi/sec), representative of the maximum predicted values (2760 psia and 240 psi/sec), for FSAR events in the PWR units listed in Table 1-1 •

To address the transition and liquid conditions resulting from FSAR events, one transition* and one {1) liquid test were perfonned. These tests {see Table 4-7, tests 931a** and 932) were perfonned at peak pressures {2578 and 2520 psi a) and pressurization rates (2.5 and 3.0 psi/sec) representative of those expected (2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test (-656°F) was representative of the upper end of the expected range (554 to 672°F). The temperature for the liquid test {463°F) envelops the lower end of the predicted liquid temperature range.


**A subsequent actuation occurred during this test (931b); however, since data for this test is incomplete, it is not discussed herein.

4-24

Based on the above, the inlet fluid conditions {state, pressure, and temperature) under which the Crosby 6M6 (Loop Seal Internals) safety valve was tested utilizing "reference" ring positions are considered adequate to represent those predicted for FSAR events in all PWR units listed in Table 1-1.


As discussed in section 2.304 1 it f s considered important to demonstrate safety valve operation at back pressures at least as high as those expected in service. Full pressure steam and loop seal tests {representative of expected FSAR events for those units listed in Table 1-1) were perfonned on the Crosby 6M6 {Loop Seal Internals) valve with •reference" ring positions with valve outlet pressures as high as 710 psiae


The maxil1lllll quasi-steady bending moment imposed on the CFasby 6M6 valve while it

was in the opening.or closing process was 298s750 in~lb {see reference 49

Table 3.5.lob, test 908)e


INLET FLUID CONDITIONS STATE/PRESSURE/TEM'ERATURE

To address possible transition and liquid eonditions resulting from Extended H1gh

Pressure Injection events, one {l) transition and one (l) liquid test were perfonnedo These tests (see Table 4-7 8 tests 93la and 932) were performed at peak pressures (2578 and 2520 psia} and pressurization rates (2.5 and 3g0 psi/sec) representative of those expected (2515 psia and 0~65 psi/see) for such eventsQ The range of liquid temperatures under which these tests were performed (463 to -656°F) 1 is Fepresentative of the maximum predicted range (400 to 650°.F). No additional testing was perforined on this valve design at lower liquid temperatures due to the expectation that the valve's perfonnance would be similar to that observed during testing performed undeF 463°F water flow conditionso

Based on the above, sufficient testing was perfonned to assess the operability of the Crosby 6M6 (Loop Seal Internals) safety valve over a range of fluid inlet conditions representative of those expected for Extended High Pressure Injection events in all PWR units listed in Table 1-lo

4-25

•

BACK PRESSURE DEVELOPED .

During transition and liquid testing of the Crosby 6M6 (Lo.op Seal Internals) safety valve, the same back pressure orifice was used in the discharge piping as was used during a full pressure steam test in which a back pressure of 710 psia

• was attained. Therefore, the back pressures developed for these liquid tests correspond to those expected under similar flow conditions in PWR units with expected steam back pressures in the 710 psia range.


See section 4.7.1.

4-26

.i. 8

N .....

Table 4-7

~As TESTED~ COMBUSTION ENG!NEERHNG MATRIX Of TESTS PERFORMED ON THE CROSBY 6M6 (LOOP SEAL HMTERNALS) SAFETY VALVE WITH mREFERENCE~ RING POSITIONS(A)

TEST TEST INLET Pl!PING · CONDHTiONS JUST PRIOR TO NO. TYPE ~ONFHGURAYiON(2) VALVE OPENING TRANSIENT COND!YIONS

P£AI? Ai VALVE KNLEi IN PEAK PEAK BENDING

PRESS ACCUr-tJlATOR TANK(l) BACK MOMENT FLUID TE!-P PRESS RATE TEf'f ?RESS PRESSURE INOUCED(4) ~ PSIA PSI/SEC FLUID~ PSUA PSiA IN-LB

929 LOOP SEAL 0061111 WATER 90 2600. 319 STEAM ~7) 2726 710 179.250 93h LS-TRANSITION Wl,\TER 117 2570 2e5 STEAM/ 1)

725(5) 93tb(6)wATER

WATER 2578 2038150 WATER (1) (9) (9) WATER (7) (9) (9)(5) (8)

932 WATER WATER 463 2501 JoO WATER 515 2520 650 1071)550 1406 LOOP SEAL WATER 141 2530 325

STEAM n 2703 250 2860800 1411 STEAM STEAM (1) 2410 300 STEAM 1 2664 245 239.000 1415 LOOP SEAL WATER 290 ·2555 360 STEAM 1 2760 255 . 268.875 1419 LOOP SEAL • STEAM 350 2464 360 STEAM (1 2675 245 256,925

(1) The r1ng positions for these tests were: For tests 929. 931ae b~ and 932: upper (-71, relative to bottom of dtsc ~ing) 0 lower (-18) and for tests 1406~ 1411, 1415~· and 1419: (-77. relative to bottom of disc ring), lower (-18)e The change fn upper ring position from -11 to -77 would be expected to result 1n a s11ght ·tncrease tn valve stabtlity0 ~owever. all tests s~own above are sttll considered "ref erence0 testse ·

(231 See Table 4-2 for conffguratton detatlse

( Valve.inlet pressures slightly lower due to frictional pressure drop tn tnlet ptptnge (4 -LateW'-(5) Same BP orf ffce used is in test 929e (67) Subsequent actuation on water du~1ng test 931G

( ) Test tn1ttated wtth saturated steam or water ~omtnally at 2300 psta (t.e., at a temperature of -656°f). (8) Not evaluatedo · (9) Not avaf 1ab1eo

(

~ .

l-

I

;..

! i I

I ~ I· I

•

•

4.8 CROSBY 6N8 (STEAM INTERNALS)

Table 4-8 presents the conditions under which the Crosby 6N8 (Steam Internals) safety valve was tested utilizing 11 reference 11 ring positions. These tests were perfonned only with a long inlet piping configuration since the operability of such valves with a shorter inlet is expected to be as good or better than that observed for the tested configuration. The following sections provide specific justification that for each event type considered {FSAR and Extended High Pressure Injection), the conditions under which this valve design was tested are representative of the limiting fluid inlet conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability (back pressure and bending moment) imposed on this valve design are discussed~

4.8.1 FSAR EVENTS


Two (2) steam tests (Table 4-8, tests 1207 and 1208) were performed on the Crosby 6N8 (Steam Internals)· safety valve with 11 reference11 ring positions at peak pressures (2674 and 2640 psia) and pressurization rates {317 and 325 psi/sec), representative of the maximum predicted values (2760 psia and 240 psi/sec) for FSAR events in the PWR units listed in Table 1-1 •

To address the transition and liquid conditions resulting from FSAR events, one (1) transition and two (2) liquid tests were performed. These tests (see Table 4-8, tests 1209d, 1211, and 1213) were performed at peak pressures (2420» 2450, and 2605 psia) and pressurization rates (5.3, 4.6, and 3.1 psi/sec), representative of those expected (2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test (-656°F) was representative of the upper end of the expected range (554 to 672°F). The liquid temperatures for the liquid tests (621 and 536°F} were representative of the full range of predicted liquid temperatures.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Crosby 6N8 (Steam Internals} safety valve was tested utilizing "reference" ring positions are considered adequate to represent those predicted for FSAR events in all PWR units listed in Table 1-1.

4-28

BACl< PRESSURE DEVELOP.ED

As discussed in section 2.3.4, it is considered important to demonstrate safety valve operation at back pressures at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table l-l) were perforined on the Crosby 6N8 (Steam Internals) valve with •referencen ring positions with valve outlet pressures as high as 560 psia.


The maxilllJlfi quasi-steady bending moment imposed on the Crosby 6N8 valve while it was in the opening or closing process was 682,500 in-lb (see reference 4, Table 3.6.1.b, test 1203).

4.8e2 EXTENDED HlGH PRESSURE INJECTION EVENTS


To addFess .possible tFansition and liquid conditions resulting from Extended High Pressure Injection events, one (1) transition and two (2) liquid tests were peFm fonnede These tests (see Table 4-8, tests 1209d, 1211, and 1213) were performed at peak pressures (2480, 2450, and 2605 psia) and pressurization rates (5o3, 5.0, and 3.1 psi/sec), representative of those expected (2515 psia and 0=65 psi/sec) for such events. The range af liquid temperatures under which these tests were performed (621 to 536°F) is representative of the uppeF end of the rnaximum predicted range (400 to 650@F)Q

No additional testing was performed an this valve design at lower liquid temper~ atures due to.the expectation that the valve's perfonnance would be similar to that observed during testing perforined under 536°F liquid flow conditions.

Based on the above, sufficient testing was performed to assess the operability of the Crosby 6N8 (Steam Internals) safety valve over a range of fluid inlet cofi

ditions representative of those expected for Extended High Pressure Injection events in all PWR units listed in Table l~lG


During transition and liquid testing of the Crosby 6N8 (Steam Internals) safety· valve, the same back pressure orifice was used in the discharge piping as was used during a full pressure steam test in which a back pressure of 560 psia was attained. TI'lerefore, the back pressures developed for these liquid tests

4-29 •

correspond to those expected under similar flow conditions in PWR units with expected steam back pressures in the 560 psia range.


• See section 4.8.1 •

•

• 4-30

~ I w ....

TEST NO.

TEST TYPE

Table 4-8

9 AS TESTED" COMBUSTION ENGINEERING MATRIX OF TESTS PERFORMED ON THE CROSBY 6N8 (STEAM INTERNALS) SAFETY VALVE WiTH 01REFERENCE11 RING POSITIONS(l)

iNLEi PIPiNG CONDKTIONS JUST PRIOR TO CONFIGURATION(2) VALVE OPENING TRANSIENT CONDXTIONS

~~~~~~~~~~~~--- PEAK Al VALVE INLET IN PEAK BENOiNG

BACK MOMENT PRESS ACCUrtJLATOR FLUID TEtt> PRESS RATE fEf'F

PEAK TANK(l) PRESS PSIA

PRESSURE INOUCEo(4) °F PSiA PSi/SEC FLUID @f PSIA IN-LB

1207 STEAM 1208 STEAM i209a TRANSRTION

1209b

1209c

1209d

1211 WATER 1213 WATER

"H" STEAM 171 2484 317 STEAM 1 . 2445 325 STEAM 7 2466 2.6

STEAM (7) (8} 5al

STEAM (7) (8) 5.0

STEAM (7) (8) . 5.3

WATER 621 2450 4.6 WATER 536 2526 3.1

STEAM (7! STEAM fl STEAWfl' WATER STEAW WATER (1) STEAJ.V WATER fl) STEAfV WATER (7) WATER 635 WATER 549

(1) The ring postt1ons for these tests were: upper (-40), lower (-18). (2) See Table 4-2 for configuration deta11sc

2674 2640

2466

2455

2480

2420 2450 2605

560 200

399(5)

402

420

570(5)

~:~mi

655 9 200 4130200

5180700

591,500 51811700

(3) Valve 1n1et pressures slightly lower due to frictional pressure drop tn inlet p1p1ng. (4) The highest bending moment imposed on the Crosby 6N8 valve while the valve was tn the opentng or

c1ostng process was 682.500 in-lb (see refe~ence 41 Table 3a6.l.b~ test 1203)e (5) Same BP orifice used as in test 1207. (6) Data oscillatory due to valve chattere (87) Test ~n1t1ated with saturated steam or water nominally at 2300 ps1a (1.e •• at a temperature of -656°f).

( ) Not available. .

••

4.9 CROSBY 6N8 (LOOP SEAL INTERNALS)

The material properties of the internals of the PWR safety valve designs selected for testing are considered important from the viewpoint of assessing valve seat leakage but are not expected to affect the operability of the valve. Therefore, testing performed on the Crosby 6N8 with steam internals safety valve -(see section 4.8), when combined with the loop seal tests performed on the Crosby 3K6 and 6M6 safety valves, provides an adequate basis to judge the operability of the Crosby 6N8 with loop seal internals safety valve over the full range of expected inlet fluid conditions (see Table 4-1} for the PWR units listed in Table 1-1.

4-32

•

4.10 TARGET ROCK' 69C

Table 4-9 presents the conditions under which the Target Rock 69C safety valve was tested. These tests were perfonned with a single inlet piping configuration representative of the inlet piping utilized in the only PWR with this valve design (Beaver Valley 1). The following sections provide specific justification that for each event type considered {FSAR and Extended High Pressure Injection) the conditions under which this valve design was tested are representative of the limiting inlet fluid conditions presented in Table 4-1. In addition, the discharge piping effects on valve operability {back pressure and bending moment), imposed on this valve design are discussed.

4.10.1 FSAR EVENTS


Two (2) steam and one loop seal test (Table 4-9, tests 722, 723 and 706) were perfonned on the Target Rock 69C safety valve at peak pressures (2678, 2674, and 27i3 psia) and pressurization rates (311, 307, and 300 psi/sec) representative of the maximum predicted values (2760 psia and 240 psi/sec) for FSAR events in the PWR units listed in Table 1-1.

To address the transition and liquid conditions resulting from FSAR events, one (1) transition and nine (9) water tests were perfonned. These tests (see Table 4-9, tests 709b, 712, and 714a through h) were perfonned at peak pressures ranging from 2390 to 2486 psi a and pressurization rates ranging from 2.2 to 8.-0 psi/sec which are representative of those expected {2575 psia and 0-12.2 psi/sec) for such events. The liquid temperature for the transition test {-656°F) was representative of the upper end of the expected range {554 to 672°F).

' '

Tests 712 and 714a were perfonned with 1 ow ·temperature (118 and 97°F) water at the valve inlet followed by 613 and 568°F water, respectively. These tests applied the maximum expected thennal shock to the valve while demonstrating its ability to operate over the full range of expected temperatures (554 to 672°F) for FSAR events. Seven additional actuations were perfonned (tests 714b through h) on liquid temperatures ranging from 510 to 550°F confinning the valve's operation at the lower end of the expected liquid temperature range.

Based on the above, the inlet fluid conditions (state, pressure, and temperature) under which the Target Rock 69C valve was tested are considered adequate to represent those predicted for FSAR events in all PWR units listed in Table 1-1 •

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As discussed in section 2.3.4, it is considered important to demonstrate safety valve operation at back pressures at least as high as those expected in service. Full pressure steam tests (representative of expected FSAR events for those units listed in Table 1-1) were perfonned on the Target Rock 69C with valve outlet pressures as high as 482 ps1a.


The max111llm quasi-steady bending manent imposed on the Target Rock 69C valve while it was in the opening or closing process was 258,750 in-lb (see Table 4-9, test 709a).

4ol0o2 EXTENDED HIGH PRESSURE INJECTION EVENTS


To address possible transition and liquid conditiqns resulting from Extended High Pressure Injection events» the transition and liquid tests discussed 1n sec-tion 4ol0.l adequately represent the peak pressure, pressurization rate range. and mid to upper range of liquid temperature predicted for such events (2515 psia, 0-65 psi/sec, and 400 to 650°F). To cover the lower end of the expected liquid temperature range, two additional liquid tests were perfonned {see Table 4-9s tests 717 and 719) with peak pressures (2490 and 2487 psia} and pressurization rates {206 and 0.7 psi/sec) in the range of those predicted for such eventso The liquid temperature at the valve inlet for these tests was very low (82-83°F) and this cold water was followed during the actuation event by 410 to 397°F water. These tests applied the maximum expected thennal shock to the valve while providing a basis for assessing valve operability on a water temperature (-400°F) representative of the low end of the predicted range.

Based on the above, the inlet fluid conditions under which the Target Rock 69C valve was tested, adequately represent those expected during Extended High Pressure Injection events in all PWR units listed in Table l•lc

BACK PRESSURE DEVELOPED ·

During transition and liquid testing of the Target Rock 69C safety valve, the same back pressure orifice was used in the discharge piping as was used during full pressure steam tests in which back pressures ranging from 270 to 482 psia were attained. Therefore, the back pressures developed for these liquid tests

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•

correspond to those expected under similar flow conditions in·· PWR units with expected steam back pressures in the 270 to 482 psia range.



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Table 4-9

wAS TESTED00 COMBUSTION ENGINEERING TEST MATRIX fOR THE TARGET ROCK 69C ; -

SAFETY VALVE

TEST lESi INU:Y IPRP!NG CONDITIONS JUST PRIOR TO NO. TYPE CONFIGURATION(l) VAUIE OPENING TRANSIENT CONDITIONS

PEAK AY VAL VE H NUET . IN PEAK PEAK BENDiNG

PRESS ACCUKJLATOR TANK(l) BACK MOMENT !

FLUID TErt> PRESS RATE f Ef'Ui PRESS PRESSURE INDUCED i

Of PSIA PSI/SEC FLUID Of PSIA PSIA IN-LB I I

703aC2ltoOP SEAL -1 le WATER 90 2543 2.7 SffAMn 2543 380 1438750 i i.

706 LOOP SEAL WATER 93 2610 300 STEAM 5 2713 482 230,000 ' ' 709a LOOP SEAL WATER 97 2508 2.0 STEAM 5 2508 270 258,750 709b TRANSITION STEAM (5) 2474 3.4 STEAM/ 5

6171:1

.,., WATER 2474 (6) I 712 WATER WATER 118 2486 2e8 WATER 613 2486 80,500 w 500 4 0\ 714a WATER WATER 97 2464 2a2 WATER 568 2464 87 201,250

714b WATER WATER 540 2433 4.2 WATER (71 2433 90

m 714c WATER WATER 510 2466 5a0 WATER 1 2466 275 -714d WATER WATER 530 2450 JeO . WATER h 2450 380 714e WATER WATER 540 2424 8a0 WATER n 2424 405

II 714f WATER WATER 542 24U 7o5 WATER 7 2411 420 714g WATER WATER 550 2396 601 WATER 1 2396 410 714h WATER WATER 541. 2390 .2 .. 1 WATER 1 2390 380(4~ 717 WATER WATER 82 2490 2.6 WATER 410 '2490 215(4 719 WATER WATER 83 2487 .. 1 WATER 397 2487 146 7 .,750 722 STEAM STEAM (5~ 2612 311 STEAM ~5~ 2678 430 57.500 723 STEAM .a. STEAM (5 2630 307 STEAM 5 2674 63 57 .,500

(1) See Table 4~2 for conftguratton details .. (2) Several subsequent actuat1ons on steam occurred during test 703 (s~e reference 4. Table 3 .. 7.1.b).

Conditions are s~own ~piy for the ftrst actuation on a loop $ea1. 1.eo 1 test 703a. (3) Valve inlet pressures s11ghtly lower due to frfct1ona1 pressure drop tn tnlet p1ptng.

~:i Same BP orifice used as in tests 703a~ 706 0 and 722. Test initiated wtth saturated steam or wate~ nominally at 2300 psta (1.e., at a temperature of -656°F).

(6) Not eva"Buated. (7) Not available.

\

•

Section 5

REFERENCES

1. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in B&W 177FA and 205FA Plants, March 1982.

2. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Combustion Engineering Plants, March 1982.

3. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse Plants, January 1982.

4. EPRI PWR Safety and Relief Valve Test Program Safety and Relief Valve Test Report, April 1982.

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'EPRI PWR Safety & Relief Valve Test Program:Test ... · NOTICE This report was prepared by the Electric Power Research Institute, Ineo (EPRI). Neither EPRI, members of EPRl, nor

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