Boiling-Water Reactor Internals Aging- Degradation Study · NRC FIN B0828 Under Contract No. DE-ACO5-840R21400. Abstract This report documents the results of a study on the effects

NUREG/CR-5754ORNL/TM-11876

Boiling-Water ReactorInternals Aging-Degradation Study

Phase 1

Prepared by

Prepared byK. H. Luk

Oak Ridge National Laboratory

Prepared forU.S. Nuclear Regulatory Commission

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NUREG/CR-5754ORNIJTM-11876RV

Boiling-Water ReactorInternals AgingDegradation Study

Phase 1

Manuscript Completed: August 1993Date Published: September 1993

Prepared byK. H. Luk

Oak Ridge National LaboratoryOperated by Martin Marietta Energy Systems, Inc.

Oak Ridge National LaboratoryOak Ridge, TN 37831-6285

Prepared forDivision of EngineeringOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001NRC FIN B0828Under Contract No. DE-ACO5-840R21400

Abstract

This report documents the results of a study on the effectsof aging degradation on 25 selected Boiling-Water Reactor(BWR) internal components. The operating environmentinside a BWR pressure vessel produces stressors that couldlead to the development of aging-related degradationmechanisms. A data base containing aging-related failureinformation for the selected internal components isestablished using data from Licensee Event Reports.Results of the failure information survey identified twomajor aging-related degradation mechanisms for reactorinternals: stress corrosion cracking (SCC) and fatigue.SCC includes intergranular SCC and irradiation-assistedSCC (IASCC).

Strategies for controlling and managing aging degradationsare based on understanding the relationship betweenstressors and the associated aging-related degradationmechanisms. The implementation of a plant HydrogenWater Chemistry (HWC) program is considered to be a

promising method for controlling SCC, which is the moreprevalent problem for BWRs. Flow-induced vibration(FMy) is the major cause of fatigue problems in BWRinternals. FIV problems are resolved either by eliminatingthe excitation sources or by detuning the structure frominput excitations. Questions remain concerning theeffectiveness of HWC in mitigating SCC (includingIASCC) in internals and in the assessment of high-cyclefatigue in a corrosive environment.

Vibration monitoring, based on neutron noisemeasurements and trending studies, is an inspectionmethod that can provide early failure detection capabilityand can improve the effectiveness of current plant in-service inspection programs. However, the large water gapand the lack of existing ex-core neutron flux monitors mayhinder the use of neutron noise vibration measurements inBWRs.

iii iii ~~~~~~NUREG/CR-5754

Contents

Page

Abstract ......... . . . . . . . .iii

List of Flgures ........ ;.vi.

List of Tables..ix

Acknowledgments..xiA ckn wled men. .... .............................................................. ......................... x

Summary .Xi..

1 Introduction .1

References1...............................................................................................................................................................1

2 BWR Internal Components .3

2.1 Functions of BWR Internals .32.2 Component.Descrip..on.4References..17

3 P..ma.yStressors 19

3.1 Applied.Loadings.1932 Environmental Stressors .193.3 Manufacturing Stressors .20References..20

4 Aging-Related Degradation Mechanisms .21

4.1 Stress Corrosion Cracking .214.2 Fatigue.224.3 Embrittlement.24

4.3.1 Radiation Embrittlement .244.3.2 Thermal Embrittlement .25

4A. Erosion.254.5 Creep and Stress Relaxation .25References............................................................................................................................................................... 26

5 Survey of Aging-Related Failures .29

5.1 ISI Program for Reactor Internals .295.2 Reported Failure Information Survey .295.3 Internal Component Failure Information Survey Results .30

5.3.1 Reported SCC Failures .305.3.2 Reported Fatigue Failures .31

References.33

.v NUREG/CR-5754

List of Tables

Table

4.1 BWR internals susceptible to SCC

4.2 BWR internals susceptible to fatigue .

4.3 BWR internals susceptible to embrittlement .

4.4 BWR internals susceptible to erosion .

4.5 BWR internals susceptible to radiation-induced creep and stress relaxation .

4.6 BWR internals and potential aging-related degradation mechanisms.

5.1 BWR internals with reported SCC .

5.2 BWR internals with reported fatigue failures .

Page

23

24

25

25

26

27

32

33

ix, ix ~~~~~~NUREG/CR-5754

Acknowledgments

The guidance and direction provided by D. M. Eissenberg,D. A. Casada, W. S. Farmer, and R. D. Cheverton in theconduct of this study are greatly appreciated. The authoralso wishes to thank A. E. Cross, L. J. Luttrell, and W. P.

Poore III for their help in obtaining the component failuredata for BWR internals. he assistance of C. C. Southmaydand E. W. Carver in preparing this report is gratefullyacknowledged.

xi xi ~~~~~~NUJREG/CR-5754

Summary

Reactor internals are located inside the pressure vessel, andthe operating environment is favorable to the developmentof time-dependent or aging-related degradation mecha-nisms. The main objective of this study is to assess theeffects of aging degradations on Boiling-Water Reactor(BWR) internal components.

Twenty-five BWR internals have been selected for theaging study; they serve five basic functions: (1) balance-of-plant steam quality control, (2) core support, (3) reactorcoolant flow control and core heat transfer enhancement,(4) housings for in-core instruments and specimen samples,and (5) core power shaping and reactivity control. Fuelassemblies, control rods, and in-core monitors are notincluded in the scope of the present study.

Stressors are conditions that can initiate and sustain thegrowth of aging degradations. Reactor internals aresubjected to stressors associated with applied loadings(thermal and mechanical), the reactor cooling water, theexposure to fast (E > I MeV) neutron fluxes, andmanufacturing processes. Flow-generated oscillatoryhydrodynamic forces and temperature fluctuations aremajor applied load stressors. The reactor cooling water cancreate an environment that is conducive to the developmentof stress corrosion cracking (SCC). Cast austenitic stainlesssteels ca l become embrittled by prolonged exposure tohigh temperatures. Long-term exposure to fast neutronfluxes can lead to changes in the mechanical properties of amaterial, which can become susceptible to brittle fractureand SCC. Chromium depletion in grain boundaries ofaustentic stainless steels and residual stresses in weld heataffected zones (HAZs) are primary stressors associatedwith manufacturing processes.

Potential aging-related degradation mechanisms for BWRinternals include SCC, fatigue, erosion, embrittlement,creep, and stress relaxation. These aging mechanismsdevelop at different rates and may not be equally importantin the design life of a reactor. Results of a componentfailure information survey identified SCC and fatigues asthe two major aging-related degradation mechanisms forBWR internals.

Major reported failures of BWR internals include cracks injet pump supports caused by SCC and fatigue, SCC in corespray spargers, fatigue cracks in feedwater spargers, andSCC (in some cases irradiation assisted) in in-core monitordry tubes. Some of these failures have led to lengthy repairtimes, but there is no indication that these failurescompromised the safety of reactor operations.

Aging degradations, if left unmitigated, will eventuallycause a failure in an affected component. It is essential tocontrol or eliminate stressors associated with the majoraging-related degradation mechanisms. The developmentof SCC requires three conditions: (1) a susceptiblematerial, (2) a corrosive environment, and (3) the presenceof tensile stresses. The elimination of any one of the threeconditions will reduce the likelihood of the development ofSCC. The effects of the welding-related chromiumdepletion process, which can make stainless steelcomponents susceptible to SCC, can be reduced by the useof low-carbon-content stainless steels. The tensile stresslevel in a structural component can be lowered by the useof larger components or by reducing the magnitude of theapplied loads. Heat sink welding methods can control thelevel of residual stresses in a weld HAZ. The implementa-tion of a plant Hydrogen Water Chemistry (HWC) programis intended to reduce the corrosiveness of the reactorcooling water by lowering its dissolved oxygen content.HWC is effective in mitigating SCC in the reactor recircu-lating water piping system, but there are insufficient data toassess its effectiveness in reactor internals. A majordisadvantage of the HWC program is the increase inradiation level in the power-generating areas. Irradiation-assisted SCC (IASCC) is not well understood, and manyactive research projects are trying to gain more insight andunderstanding into this phenomenon.

Flow-induced vibration is the major cause of fatigueproblems in reactor internals. Low-cycle fatigue is causedby large-amplitude vibrations and can be prevented bydetuning the system's natural frequency from the dominantflow-generated excitation frequency. Large-amplitudevibrations are usually detected and corrected during reactorpreoperational flow testings. Reactor internals aresusceptible to high-cycle fatigue caused by small-amplitude vibrations, which are much more difficult tocontrol and eliminate. The synergistic effects of high-cyclefatigue and a corrosive medium are not well quantified atthe present time. High-cycle fatigue is an active aging-related degradation mechanism for reactor internals.

In the presence of active aging-related degradationmechanisms, it is essential to maintain a vigorousinspection program to ensure the structural integrity ofreactor internals. The plant in-service inspection (ISI)program calls for the visual inspections of accessible areasof internals during refueling outages. The limitations of thevisual inspection method are well known, and variousalternate inspection methods, such as ultrasonic and eddy-current inspections, have been tried on an experimentalbasis. Reactors licensed after 1978 are equipped with loosepart monitoring systems (LPMSs), which can indicate the

xiii xiii ~~~~~~NUREGICR-5754

Summary

presence of loose parts in the reactor primary system.Appropriate actions taken by plant operators can limitfurther damage to other reactor components and systems.LPMSs do not possess the capability of indicating thecurrent status of flaws that may exist in a component.Neutron noise vibration measurements and trending studiescan provide the information that can be used to evaluatethe current status of flaws that may exist in selected core

internal components. However, these practices have notbeen fully exploited by the domestic utility industry.Visual inspection, supplemented by ultrasonic and eddy-current inspection, remains the major method forinspecting reactor internals. The implementation of amonitoring system with early failure detection capabilitywill further enhance the safety and efficiency of reactoroperations.

NUREG/CR-5754xi xiv

1 Introduction

The effects of aging or time-dependent degradation onsystems, structures, and components of a commercialnuclear power plant are of concern to the U.S. NuclearRegulatory Commission (NRC). The Office of NuclearRegulatory Research (RES) has initiated and sponsored aNuclear Plant Aging Research (NPAR) Program.1 Themain objective of the program is to study aging-relateddegradation mechanisms and their effects on major reactorsystems, structures, and components. The NPAR Programalso includes evaluation of the effectiveness of inspection,monitoring, and maintenance methods for managing agingeffects. The basic NPAR approach is to identify majorreactor systems, structures, and components that are sus-ceptible to the effects of aging degradation. Detailed agingstudies are then conducted for these systems, structures,and components. One of the Oak Ridge National Labora-tory's (ORNL's) assignments is to evaluate the effects ofaging degradation on reactor intemals.

Reactor internals are components located inside the reactorpressure vessel (thus, the term "internal'). The componentsperform several different functions. Many internals areparts of the core support structure, while other componentsdirect the coolant flow through the vessel and control theheat transfer in the core. Housings for in-core monitors andspecimen samples are also considered as reactor internals.Internals for Boiling-Water Reactors (BWRs) andPressurized-Water Reactors (PWRs) are different in designand construction, and they will be treated in separate stud-ies. This Phase 1 report will concentrate on the effects ofaging degradation on BWR internals.

Thirty-seven BWRs are licensed for commercial operationsin the United States.2 Using the commercial operationstarting date as the reference for counting reactor ages, 7reactors (or -19% of the total) are over 20 years old; 16reactors (or 43%) are between 10 and 20 years old; and 14reactors (or 38%) are <10 years old. There is a total of-522 reactor-years of BWR operations; the operating his-tories of these reactors provide useful information forstudying aging effects in selected reactor components.

components. The second step is to identify stressors thatare present in the operating environment of the selectedcomponents. The third step is to establish the relationshipbetween the stressors and potential aging-related degrada-tion mechanisms. The final step is the identification ofmajor aging-related degradation mechanisms based on areview of the operating history of the reactors and reportedcomponent failure information. The understanding of therelationship between major stressors and associated aging-related degradation mechanisms also provides the basis forformulating strategies for controlling and managing agingeffects.

Twenty-five BWR vessel internals selected for the presentstudy are identified in Chap. 2, which will also provide abrief description of the function, design, and constructionof each component Major stressors for these selected reac-tor internals are discussed in Chap. 3, and potential aging-related degradation mechanisms associated with the stres-sors are identified in Chap. 4. Chapter 5 summarizes thereported aging-related BWR internal failure information.The NPAR Program also addresses issues involving theinspection and maintenance methods used to control andmanage aging effects. The effectiveness of current in-service inspection (ISI) programs and strategies for manag-ing aging degradations are discussed in Chap. 6. Importantresults of this Phase 1 aging assessment of BWR internalsare summarized in Chap. 7.

References

1. J. P. Vora, "Nuclear Plant Aging Research (NPAR)Program Plan," USNRC Report NUREG-1 144,Rev. 1, September 1987.*

2. M. D. Muhlheim and E. G. Silver, "Operating U. S.Power Reactors:' Nucl. Saf. 32(3) (July-Septemberl9 91).t

Available for purchase from National Technical Information Service,Springfield, VA 22161.

tAvailable in public technical libraries.

Aging assessment is a multiple-step process. The first stepis the identification and description of the selected internal

1 1 ~~~~~~~NUREG/CR-5754

2 BWR Internal Components

This study considers the effects of aging degradations onreactor internal components in BWRs designed by theGeneral Electric Company (GE). It includes reactor designversions from BWR-2 to BWR-6. The only significant dif-ference between these reactors is that jet pumps are usedbeginning with BWR-3 and in all subsequent BWRdesigns. Following are the 25 BWR internals selected forthis study:

1. steam dryer2. steam separator3. steam separator support ring4. shroud head5. shroud head bolts6. top guide7. core shroud8. core plate9. orificed fuel supports (OFSs)

10. shroud support including access hole covers11. core spray internal piping12. core spray sparger13. feedwater sparger14. jet pump assembly15. in-core neutron flux monitor housings16. in-core neutron flux monitor guide tubes17. in-core neutron flux monitor dry tubes18. control rod drive (CRD) housing19. jet pump sensing line20. neutron source holder21. control blades22. vessel head cooling spray nozzle23. control rod guide tubes24. surveillance sample holders25. differential pressure and liquid control line.

Most of the 25 selected components are commonly identi-fied as "internals" in the Final Safety Analysis Report(FSAR) for a BWR plant; the three exceptions are the in-core neutron flux monitor housings, the CRD housing, andin-core neutron flux monitor dry tubes. The CRD and in-core neutron flux monitor housings are welded to the reac-tor vessel and are usually considered parts of the pressurevessel. Dry tubes can be considered as a part of the in-coreneutron flux monitor. These components are included as*internals because they perform a basic internal function(i.e., providing housings to in-core instruments).

The fuel assemblies, control rods, and in-core monitors arereactor components also located inside the reactor vessel.They perform unique functions that are different fromfunctions of reactor internals. Fuel assemblies, controlrods, and in-core monitors will not be included in the scopeof the present study.

A simplified sketch of the arrangement of major BWRinternal components with jet pumps is shown inFig. 2.1(a), and the arrangement of internals in a BWRwithout jet pumps is shown in Fig. 2.1(b).

2.1 Functions of BWR Internals

The 25 selected BWR internals perform a variety of func-tions. They can be grouped into five general areas:(1) balance-of-plant (BOP) steam quality control, (2) coresupport structures, (3) reactor coolant flow control and coreheat transfer enhancement, (4) housings for in-core instru-ments and specimen samples, and (5) core power shapingand reactivity control.

Two of the selected intemals, the steam dryer and steamseparators, are used for BOP steam quality control. Eightof the 25 components provide structural supports to thecore and other internals: steam dryer support ring, shroudhead, shroud head bolts, top guide, core shroud, core plate,OFSs, and access hole cover. The next largest group ofinternals are housings for in-core instruments and specimensamples: in-core neutron flux monitor housings, in-coreneutron flux monitor guide tubes, in-core neutron fluxmonitor dry tubes, CRD housings, neutron source holder,control rod guide tubes, jet pump sensing line, and surveil-

-lance sample holders. Six components are used for coolantflow control and heat transfer enhancement: core sprayinternal piping, core spray sparger, feedwater sparger, ves-sel head cooling spray nozzle, differential pressure and liq-uid control line, and the jet pump assembly. Control bladesand jet pumps are used for core power shaping and reactiv-ity control.

The majority of the BWR internals are made of type 304stainless steel. They are designed in accordance with therules and regulations of Sect. III of the American Society ofMechanical Engineers Boiler and Pressure Vessel (ASMEB&PV) Code.1 Specifically, internal components that areparts of the core support system and those that are used forreactor coolant flow control and core heat transfer enhance-ment are treated as safety-class items and are designed tomeet the requirements as stipulated in Appendix 1 ofSect. III of the ASME B&PV Code. CRD housings, controlrod guide tubes, and in-core monitor housings that are apart of the reactor primary pressure boundary are alsodesigned to meet all Sect. III requirements. Allowablestress values from Sect. III are used as design guides forinternal components that are not safety-class items. Safety-class internal components are fabricated in accordance withrequirements of Sect. III, Subsection NG; other internalsare fabricated to meet requirements of Sect. IX of the

3 NUREG/CR-5754

BWR

ORNL-DWG 91-2959A ETD

Figure 2.1(a) BWR internals with jet pumps

ASME B&PV Code.2 For reactors that were designed andbuilt before the establishment of Sect. III, analyses areperformed to ensure that calculated stress values for suchreactor components meet the intent of the Code underspecified design conditions.

The following is a brief description of the function, design,and construction of the 25 selected BWR reactor internalcomponents.

2.2 Component Description

Information on the design of the 25 selected BWR internalsis obtained from various plant FSARs and supplementedby the Electric Power Research Institute (EPRI) report onthe Monticello pilot plant life extension study.3

Reactor design is an evolving process, and many designfeatures are plant specific. Internal components included inthis report are those for a "typical" reactor designed by thevendor. The report will attempt to identify major changes

NUREGICR-57544 4

BWR

OSPNt-OMi 9~3-34 ETU

VDUEL HADSLMDL

-VESSEIL 300 n

STEAM PRESSURE

A D~~~~~~~~~fD~MA UREMNTTA

5 ~ ~~~~~~~~~~CNTRO G/L5NW

5s NUREG/CR-5754

BWR

in internal components of the different reactor design ver-sions, but no attempt will be made to identify all designchanges.

Component No. 1: Steam Dryer

Function: The steam dryer removes the excess moisturein the steam exiting from steam separators. It is not neededfor steady-state operations in the newer BWRs.-

Description: The steam dryer assembly consists of thedrying vanes and the top and bottom structural supportmembers. They form a single structural unit that ismounted above the steam separators. The entire unit issupported by brackets extending inward from the pressurevessel wall. The moisture removed by the dryer is carriedby a system of troughs and drains to the downcomer,which is the annular space between the core shroud and thepressure vessel wall. A sketch of a steam dryer unit isshown in Fig. 2.2.

Design Code: The steam dryer is designed to meet theintent of Sect. III of the ASME B&PV Code.

Material of Construction: The steam dryer is made oftype 304 stainless steel.

Component No. 2: Steam Separators

Function: The steam separators separate water dropletsfrom the steam in the steam-water mixture generated in thecore.

Description: The steam separator assembly consists of129 standpipes welded to openings in the shroud head, andan axial-flow steam separator is attached to the top of eachstandpipe. Fixed vanes are located inside the separator. Asthe steam-water mixture rises through the standpipes, avortex is generated by the spinning action imparted to themixture by the vanes. The vortex flow separates the waterfrom the steam in each of the three stages of the separator.The steam passes through the top of the separators and intothe dryer assembly. The water flows down the standpipesand into the annular space between the core shroud and thepressure vessel wall. A sketch of the steam separatorassembly is shown in Fig. 2.3.

Design Code: The steam separator is designed to meet theintent of Sect. III of the ASME B&PV Code.

Material of Construction: Standpipes and steam separa-tors are made of type 304 stainless steel.


STEAMFLOW

COLLECTING 'TROUGH-

DRAIN TUBES

STEAMFLOW

Figure 2.2 Steam dryer assembly

6NUREG/CR-5754

BWR

ORNL-DWG 91-2961 ETD head bolts. Steam separator standpipes are welded to open-ings in the shroud head. A sketch of the shroud head isshown in Fig. 2.4.

Design Code: The shroud head is designed to meet theintent of Sect. III of the ASME B&PV Code.

Material of Construction: The shroud head is made oftype 304 stainless steel.

ORNL-DWG 91-2962R ETD

RETURNING -WATER

CORE DISCHARGEPLENUM

Figure 2.3 Steam separator assembly

Component No. 3: Steam Dryer Support Ring

Function: The steam dryer support ring supports thesteam dryer assembly. The ring is attached to four supportbrackets that are welded to the pressure vessel wall.

Description: The circular support ring is'made of cold-formed steel sections. The steel section is rectangular andthe dimensions are 3 by 6 in. Some BWR plants have'steam dryer support rings made of non-cold-worked steelsections. -'

Design Code: The support ring is designed to meet theintent of Sect. III of the ASME B&PV Code.

Material of Construction: The steam dryer support ringis made of type 304 stainless steel.

Component No. 4: Shroud Head

Function: The shroud head covers the discharge plenumregion'located at the top of the core. The plenum serves asa mixing chamber for the steam-water mixture before itenters the steam separators. The shroud head also providesstructural support to the steam separators.

Description: The shroud head is a dome-shaped steelstructure attached to the core shroud top flange by shroud

Figure 2.4 Shroud head

Component No. 5: Shroud Head Bolts

Function: Shroud-head bolts fasten the shroud head to thecore shroud top flange. Typically, 36 bolts are used; how-ever, the exact number of bolts and bolt sizes can varyfrom plant to plant.

Description: A typical shroud-head bolt is 1.75 in. indiameter and 14 ft long. A nut is screwed onto one end ofthe bolt, and a tee head is welded to the other. A sleevecovers the rest of the bolt, and the base of the sleeve' isjoined to a collar that is welded to the shaft near the teehead. A part of the collar is cut out to provide space for thealignment pin window. A sketch of a shroud-head bolt isshown in Fig. 2.5.

Design Code: The shroud head bolt is designed to meetthe requirements of Sect.-Ill of the ASME B&PV Code.

Material of Construction: The bolt, the nut, and the teehead are made of Inconel 600 alloy. The sleeve and thecollar are made of type 304 stainless steel.

;7 7 ~~~~~~~NUREG/CR-5754

BWR

ORNL-DWG 91-2963

BOLT

NUT

.SLEEVE

- SLEEVE BASE

ORNL-DWG 91-2964 ETD

Figure 2.6 Top guide

WINDOWDesign Code: The top guide is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Figure 2.5 Shroud head bolt

Component No. 6: Top Guide

Function: The top guide provides lateral support andmaintains proper spacing of the upper ends of the fuelassemblies.

Description: Top guides, with the exception of those inBWR-6s, are made by welding intersecting steel beams toa circular rim. At the intersecting joints, the upper beamshave slots cut on the lower part, and the lower beams haveslots cut on the upper part. The upper and lower beams areinterlocked to form the grid. A crevice-free design, inwhich the rim and intersecting beams are machined from asingle piece of steel, is used in BWR-6s. Each squareopening formed by the intersecting beams provides lateralsupport and maintains proper spacing for four fuel assem-blies. Holes are drilled into the bottoms of the beams, andthey provide support to in-core neutron flux monitor drytubes and the start-up neutron sources. The top guide isaligned by positioning pins that fit into slots in the top ofthe core shroud. A simplified sketch of a top guide isshown in Fig. 2.6.NUREG/CR-5754

Material of Construction: The top guide is made of type304 stainless steel.

Component No. 7: Core Shroud

Function: The core shroud provides lateral restraint to thecore. The top guide and the core plate are also supportedby the core shroud. The core shroud serves as the partitionseparating the upward coolant flow from the downwardrecirculating flow in the pressure vessel.

Description: The core shroud is a segmented cylindricalsteel structure. Each segment has a different shroud diame-ter. The steam separator/shroud head assembly is bolted tothe flange of the top segment, which has the largest diame-ter. Core spray spargers are attached to the inside surfaceof the top shroud segment. The top guide is the lower -.

boundary of the top segment. The middle shroud segmentcontains the core. It is the longest segment with an inter--mediate diameter. The top guide and the core plate definethe top and bottom boundaries of the middle segment. Thebottom segment of the core shroud contains the upper por-tion of the reactor lower plenum. The bottom shroud seg-ment has the smallest diameter of the three. The bottom ofthe core shroud is welded to the shroud support.

8

BWR

For older reactors that are not equipped with jet pumps, thebottom shroud segment is in the form of a truncated cone.The top end of the cone segment is welded to the lower endof the middle shroud segment, and the bottom of the coneis welded to the reactor vessel wall.

Design Code: The core shroud is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Material of Construction: The core shroud is made oftype 304 or 304L stainless steel plates. The shell sectionsare solution heat-treated and then joined together by longi-tudinal and circumferential welds.

Component No. 8: Core Plate

Function: Perforations in the core plate provide lateralsupport and guidance to control rod guide tubes, peripheralfuel support pieces, in-core neutron flux monitor housings,and start-up neutron sources.

Description: The core plate assembly consists of a perfo-rated top plate stiffened by a circular rim and grid supportbeams. With the exception of BWR-6s, the top plate is alsosupported by tie rods. Control rod guide tubes, peripheralfuel support pieces, and the in-core neutron flux monitorhousings are inserted into the perforations of the top plate.Grid support beams are attached to the core shroud, andthey are fillet-welded at regular intervals to the top plate.Tie rods, when they are used, provide lateral support to thegrid support beams. One end of the tie rods is fillet-weldedto grid support beams through holes in the top plate, andthe other end is attached to the rim at the periphery of thetop plate. A simplified sketch of a core plate is shown inFig. 2.7.

Design Code: The core plate is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Material of Construction: The core plate assembly ismade of type 304 stainless steel.

Component No. 9: OFS Pieces

Function: OFS pieces support the weight of the fuelassemblies, as well as distributing cooling water to them.

Description: There are two types of OFS pieces. Thestandard or four-lobed OFS is a cylindrical structure withfour internal compartments and a central opening for thepositioning of a control blade. Each of the four internalcompartments also has a lobe-shaped opening on the topsurface. The standard OFS piece rests on top of a controlrod guide tube. The bottom of a fuel assembly is inserted


A A

RODSGRID BEAMS

SECTION A- A

Figure 2.7 Core plate

into one of the lobe-shaped openings, and the OFS pieceprovides lateral support and alignment to the bottom of thefuel assemblies. The weight of the fuel assembly is trans-ferred to the control rod guide tubes through the OFSpiece. The coolant flow into the fuel assemblies is regu-lated by an orfice located on the side of the lower portionof the standard OFS piece. The peripheral OFS piece is aone-opening cylindrical structure that provides lateral sup-port to only one fuel assembly. The peripheral OFS iswelded to perforations in the core plate. The orifice of theperipheral OFS, which regulates coolant flow into the fuelassemblies, is located directly below the top opening. Asimplified sketch of a standard OFS piece is shown inFig. 2.8.

Design Code: The OFS is designed to meet the require-ments of Sect. III of the ASME B&PV Code.

Material of Construction: The center or standard OFSpiece is cast from Grade CF-3 or CF-8 steel. The periph-eral OFS is made of type 304 or 304L stainless steel.

9 9 ~~~~~~~NUREG/CR-5754

BWR

ORNL-DWG 91-2966A ETD ORNL-DWG 91-2967 ETD

ACCESS HOLE COVERA

L

I

SECTION A- A

Figure 2.8 Standard OFS piece

Component No. 10: Shroud SupportIncluding Access Hole Covers

Function: The shroud support, also known as the shroudbaffle plate, carries the weight of the shroud, shroud head,steam separators, peripheral fuel assemblies, startup neu-tron sources, core plate, top guide, and jet pump diffusers.The shroud support also provides lateral support to the fuelassemblies. Jet pump diffusers penetrate the plate to injectcooling water to the inlet plenum below the core.

Description: The shroud support is an annular plate. Theouter edge of the plate is welded to the inside surface of thereactor pressure vessel, and the inner edge is welded to thebottom segment of the core shroud. The inner edge of theannular plate is also supported by columns welded to thevessel bottom head. Jet pump diffusers are attached toopenings in the shroud support plate. Two access holes,1800 apart, provide access to the jet pumps during con-struction. The access hole covers plugged the two accessopenings. The access hole cover is a circular steel plate. Itis put into the access opening and rests on a ledge near thebottom of the hole. A full penetration weld attaches thecover to the shroud support. A sketch of the access holecover is shown in Fig. 2.9.NUREG/CR-5754

Figure 2.9 Access hole cover

Design Code: The access hole cover is designed to meetthe requirements of Sect. III of the ASME B&PV Code.

Material of Construction: The access hole cover is madeof alloy 600.

Component No. 11: Core Spray Line InternalPiping

Function: The core spray line internal piping, commonlyreferred to as the core spray line, is a component of thecore spray system that supplies cooling water to the reactorfuel assemblies during a loss-of-coolant accident (LOCA).The objective is to supply and distribute sufficient coolantto the fuel assemblies so that the maximum fuel claddingtemperature of 1204TC (22001F) is not exceeded during aLOCA.

Description: The core spray line connects the externalcore spray piping to the core spray spargers. It is made of a5-in. Schedule 40 steel pipe, and the content in the pipe isin a stagnant condition during normal plant operations.Two core spray lines enter the reactor vessel through twocore spray nozzles. The two nozzles are located 1800 apartin the vessel wall. Upon entry into the reactor vessel, a'core spray line divides into two halves that are routed toopposite sides of the vessel. Along the way, the pipe linesare supported by clamps attached to the vessel wall. Thepipes then go down the downcomer and are butt-welded toone end of the tee box pipe sections. The tee box pipe sec-tions enter the top core shroud segment below the topflange. They then pass through the rim of the top guide and

4

I

10

BWR

are connected to the center of a semicircular core spraysparger. The tee box pipe section is a component of thecore spray sparger. The route of the other core spray line isidentical, except it is connected to a core spray sparger at adifferent elevation. When the core spray system is acti-vated in a LOCA, the core spray spargers inject coolingwater into the core.

Design Code: The core spray line internal piping isdesigned to meet the requirements of Sect. HII of theASME B&PV Code.

Material of Construction: The core spray line internalpiping is made of type 304 stainless steel.

Component No. 12: Core Spray Sparger

Function: The core spray sparger is a part of the CoreSpray System. It injects cooling water into the core in theevent of a LOCA. The objective is to supply and distributesufficient coolant to the fuel assemblies, so the maximumfuel cladding temperature of 12041C (22001F) is notexceeded during a LOCA.

Description: The core spray spargers consist of two circu-lar headers at different elevations and two tee box pipe sec-tions for each header. The upper header has bottom-mounted nozzles, and the lower header has upper-mountednozzles. Each circular header is composed of two 1800segments made of 3-1/2-in. Schedule 40 steel piping. Eachof the 180° header segment is supplied by a tee box pipesection. The tee box pipe section, made of a 5-in. Schedule40 steel pipe, connects the core spray spargers to the corespray line internal piping. One end of the tee box pipe sec-tion extends through the shroud wall and is butt-welded tothe core spray line. The other openings of the tee box pipesections are connected to the 1800 header segments. Thetee box pipe section is attached to the shroud by the sealrings with the attachment welded to the 5-in. pipe and theexterior surface. of the shroud. The content in a core spraysparger is in a stagnant condition during normal operations.A simplified sketch of a core spray sparger is shown inFig. 2.10.

Design Code: The core spray sparger is designed to meetthe requirements of Sect. 11 of the ASME B&PV Code.

Material of Construction: The core spray sparger ismade of type 304 stainless steel.

Component No. 13: Feedwater Sparger

Function: During normal operation, the feedwater spargerdistributes cooler feedwater to the saturated reactor recircu-lating water before it comes into contact with the vessel


-LOWER SPARGERTEE BOX

VEW A- A

Figure 2.10 Core spray sparger

wall. The mixing of the two flows produces a flow of sub-cooled water to the jet pump and recirculation pump inletsand prevents the occurrence of pump cavitation problems.The homogeneous and uniform temperature mixture willalso help to prevent the development of an asymmetricalpower distribution in the reactor core. In the event of aLOCA, the feedwater sparger becomes a component of theHigh-Pressure Coolant Injection'(HPCI) System. TheHPCI System injects cooling water through the feedwatersparger into the core to maintain an adequate reactor waterlevel in a LOCA.

Description: The feedwater sparger is a segmented circu-lar header made of 5-in. Schedule 40 curved pipe sections.The number of segments is either four or six and is plantspecific. Each segment receives water from a feedwaternozzle. The water inlet is located in the middle of eachsparger segment, and the segment is shaped to fit the con-tour of the reactor vessel wall. A thermal sleeve is weldedto the water inlet; the other end of the thermal sleeve isconnected to the safe end of the feedwater nozzle by a slip-fit joint. Each sparger is supported by the thermal sleeveand a bracket mounted to each end of the segment. The endbrackets are bolted to'vessel wall brackets. The weight of

11 11 ~~~~~~~NUREG/CR-5754

BWR

the segment is supported by the vessel wall brackets, whichalso locate the header segments away from the vessel wall.Radial differential thermal expansions of the header seg-ments are accommodated by the thermal sleeves; tangentialslots are cut into the end brackets to account for tangentialdifferential thermal expansions. The sparger segments aremounted at one elevation in the vessel. Feedwater isejected through multiple small elbow and nozzle assem-blies in an inward and downward direction. A simplifiedsketch of a feedwater sparger segment is shown inFig. 2.11.

Design Code: The feedwater sparger is designed to meetthe intent of Sect. III of the ASME B&PV Code.

Material of Construction: The feedwater sparger is madeof types 316L and 316NG stainless steel.


throat or mixing section, and a diffuser. The inlet sectionconsists of a transition piece casting, a holddown beam, aninlet elbow, and a driving nozzle. The inlet section, thesuction inlet, and the throat section form a removable unit.The diffuser is welded to an opening in the ring segment ofthe reactor vessel shroud support structure.

The riser pipe is welded to the thermal sleeve of the recir-culating water inlet nozzle; the riser is also supported bytwo brace arms welded to beams extending from pads onthe vessel wall. The connection between the throat and thediffuser is a slip-fit joint. The throat section is alsorestrained laterally by a bracket attached to the riser pipe.A metal-to-metal, spherical-to-conical seal joint is usedbetween the riser pipe and the transition piece casting. Atight contact at the seal joint is maintained by clamps thatfit under ears in the riser pipe, and a holddown beam isused to exert a downward force on a pad on top of thetransition piece casting. A simplified sketch of ajet pumpassembly is shown in Fig. 2.12.

The riser pipe receives high-pressure water from the recir-culating water inlet nozzles and delivers the flow up to atransition piece casting. The flow from the riser pipe isturned downward into the driving nozzle by the inletelbow. The high-speed nozzle exit flow entrains additional

ORNL-DWG 91-2970 EMD

Figure 2.11 Feedwater sparger

Component No. 14: Jet Pump Assembly

Function: Jet pumps provide coolant flow for forced con-vection heat transfer in the reactor core, and they also pro-vide power distribution shaping and reactivity control.They are used beginning with BWR-3 designs.

Description: Jet pump assemblies are located in twosemicircular groups in the downcomer annular regionbetween the core shroud and the reactor vessel wall. Ajetpump assembly is composed of the riser pipe (shared withan adjacent jet pump), an inlet section, a suction inlet, the

NUREG/CR-5754

-RPVRECIRCULATIONINLET NOZZLE(1 TO EACH JETPUMP RISER)

Figure 2.12 Jet pump assembly

12

BWR

ORNt-DWO 93-3404 EDcooling water through the suction inlet. The two flows mixin the throat or mixing section before discharge through thediffuser. The hydrodynamic forces acting on the jet pumpsare reacted by holddown beams placed across the top of thetransition piece casting.

Design Code: The jet pump is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Material of Construction: Most components of the jetpump are made of type 304 stainless steel. Parts that werecast, such as the transition piece casting and inlet elbows,are made of either Grade CF-3 or CF-8 steel. A portion ofthe diffuser is made of alloy 600. The holddown beam andtwo parts at the inlet of the mixing section are made ofalloy X-750.

Component No. 15: In-Core Neutron FluxMonitor Housings

Function: In-core neutron flux monitor housings providea path for the insertion of neutron flux monitors into thereactor pressure vessel. Source Range Monitors (SRMs),Intermediate Range Monitors (IRMs), Local Power RangeMonitors (LPRMs), and calibration monitors enter thereactor vessel through monitor housings.

Description:' The in-core neutron flux monitor housing isa pipe segment with a seal-ring flange at the bottom. Thepipe segment is inserted into the reactor vessel throughpenetrations in the vessel bottom head and welded to theinside surface of the bottom head. An'in-core neutron fluxmonitor guide tube is welded to the top of each housing.Either an SRMAIRM drive unit, an LPRM, or a calibrationmonitor is bolted to the seal-ring flange at the bottom. Thehousing unit is considered part of the reactor primary pres-sure boundary. The configuration of a typical in-core neu-tron flux monitor is shown in Fig. 2.13. A simplifiedsketch of an in-core neutron flux monitor housing is shownin Fig. 2.14.

Design Code: The in-core neutron flux monitor housing isdesigned to meet the requirements of Sect. III of theASME B&PV Code.

Material of Construction: The in-core neutron flux moni-tor housing is made of type 304 stainless steel.

Component No. 16: In-Core Neutron FluxMonitor Guide Tubes

Function: The in-core neutron flux monitor guide tubeprovides a path for the insertion and positioning of neutron

Figure 2.13 In-core neutron flux monitor

flux monitors into the core. IRM, SRM, LPRM, and cali-bration monitors are inserted into the core through monitorguide tubes.

Description: The in-core neutron flux monitor guide tubeis a pipe welded to the top of an in-core neutron flux moni-tor housing. The pipe segment extends from the top of theneutron flux monitor housing to the top of the core plate. Alatticework of clamps, tie bars, and spacers provide struc-tural support to the guide tube. To prevent loosening dur-ing reactor operations, bolts and clamps used in buildingthe support latticework are welded after assembly. The in-core neutron flux monitor guide tube is considered part ofthe reactor primary pressure boundary. A simplified sketchof the in-core neutron flux monitor guide tube is shown inFig. 2.14.

Design Code: The in-core neutron flux monitor guide tubeis designed to meet the requirements of Sect. Ill of theASME B&PV Code.

Material of Construction: In-core neutron flux monitorguide tubes are made of type 304 stainless steel.

NUREG/CR-575413

BWR

ORNL-OWG 93-3405 EM ORNL-DWG 91-2971 ETM

guide tube

, -

, housing

Reactor vessel,. _

-GUIDETUBE

"k"Fe Figure 2.15 In-core neutron flux monitor dry tube

Figure 2.14 In-core neutron flux monitor housing andguide tube

Component No. 17: In-Core Neutron FluxMonitor Dry Tubes

Function: In-core neutron flux monitor dry tubes houseneutron monitors inside the core. Monitors include theIRM, SRM, LPRM and calibration monitors.

Description: The dry tube is a thin-walled tubing. Thelower end of the tubing is welded to a thick-walled tubethat contains the instrument cavity. The thick-walled tub-ing is a part of the reactor primary pressure boundary. Aguide plug is welded to the top of the dry tube. An adapterunit is inserted into the dry tube through the guide plug andis pressed against a spring located inside the dry tube. Thetop of the adapter is inserted into slots in the top guide. Asimplified sketch of a typical dry tube is shown inFig. 2.15.

Cbmponent No. 18: Control Rod Drive(CRD) Housing

Function: The CRD housing provides access into thereactor pressure vessel for the CRD. It supports the CRDand the control blade. Ibe CRD housing also provides thenecessary restraint needed to counteract the forces causedby the activation of the CRD system.

Description: The CRD housing is a tubular structure. Themain portion consists of two tubes welded together. Aflange is welded to the lower end, and a cap is welded tothe upper end. The CRD is bolted to the lower flange. TheCRD housing penetrates the pressure vessel bottom headand is J-welded to the top of the bottom head-stub tube. Inthe newer BWR plants, the CRD housing unit is weldeddirectly onto the bottom head. There is a thermal sleeve inthe inside surface along the entire length of the CRD hous-ing. The CRD housing is a part of the reactor primary pres-sure boundary. A simplified sketch of a CRD housing isshown in Fig. 2.16.

Design Code: The in-core neutron flux monitor dry tube isdesigned to meet the requirements of Sect. III of theASME B&PV Code.

Material of Construction: The dry tube is made ofannealed type 304 stainless steel.

Design Code: The CRD housing is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Material of Construction: The CRD housing is made oftype 304 stainless steel.

NUREG/CR-5754 114

BWR

ORNL-DWG 91-2972A ETD Component No. 20: Neutron Source Holder

Function: The neutron source holder contains theantimony-beryllium start-up sources that are needed toprovide additional neutrons for attaining criticality in thefirst operating cycle. Start-up sources are not needed afterthe first fuel cycle.

Description: The neutron source holder is a cylindricalcapsule containing the antimony gamma sources and theberyllium sleeve. The holders are fit into a slot in the topguide and inserted into a hole in the lower core plate. Thereactor vendor has recommended that neutron source hold-ers be removed from the reactor after the first fuel cycle.

Design Code: The neutron source holder is designed tomeet the intent of Sect. III of the ASME B&PV Code.

Material of Construction: The neutron source holder ismade of type 304 stainless steel.

Component No. 21: Control Blade

Function: Control blades are used for power distributionshaping and reactivity control. Power distribution shapingis achieved by the manipulation of the pattern of neutron-absorbing rods in the control blades. The neutron-absorbing rods can also be positioned to neutralize theeffects of steam voids in the top of the core for reactivitycontrol.

Description: The major component of a control blade isthe neutron-absorbing rod assembly. Neutron-absorbingrods are made of stainless steel tubings filled with boron-carbide powder. The tubes are seal-welded with end plugsat the two ends. The rods are held in a cross-shaped patternby a steel sheath extending the full length of the rod. Thesheath is spot-welded to the top of the tubes. An upperhandle aligns the tubes and provides the necessary struc-tural rigidity at the top of the rod assembly. The lowerhandle and the velocity limiter serve the same purpose atthe bottom of the assembly. Control blades are insertedinto the core from the bottom of the pressure vessel. Asimplified sketch of a control blade is shown in Fig. 2.17.

Design Code: The control blade is designed to meet therequirements of Sect. III of the ASME B&PV Code.

Material or Construction: Control blades are made oftype 304 stainless steel.

Figure 2.16 CRD housing

Component No. 19: Jet Pump Sensing Line

Function: The jet pump sensing line provides housing forinstruments that are used to monitor the jet pump flow.

Description: The jet pump sensing line is made of a1/4-in. Schedule 40 pipe (larger pipe for BWR-6s). Thesensing line is welded to the jet pump diffuser at twosquare support brackets. The line is routed down the dif-fuser, around the jet pump, and exits the reactor pressurevessel through one of the two jet pump instrumentationnozzles. There is one sensing line for each jet pump in thereactor.

Design Code: The jet pump sensing line is designed tomeet the intent of SecL III of the ASME B&PV Code.

Material of Construction: The jet pump sensing line ismade of type 304 stainless steel.

15 15 ~~~~~~~NUREG/CR-5754

BWR

ORNL-DWG 91-2973 ETD Material of Construction: The vessel head cooling spray

UPPER nozzle is made of type 304 stainless steel.HANDLE

Component No. 23: Control Rod GuideTubes.

Function: A control rod guide tube provides lateral guid-o a ance to a control rod and vertical support for a four-lobedlo OFS piece and the four fuel assemblies surrounding a con-o \o NEUTRON trol rod.I ~NEUTRON

ABSRODS Description: The control rod guide tube is a cylindricalROdS 1 tube extending from the top of the CRD housing up

o ~~~~~~~~through holes in the core plate. The bottom of the guideoa tube is supported by the CRD housing. A thermal sleeve isA o inserted into the CRD housing from below and then rotatedSHEATH to lock the control rod guide tube in place. The top cf the

SHEATH 9- tPguide tube goes through a hole in the core plate, and afour-lobed OFS piece is inserted into the top opening of theguide tube. A simplified sketch of a control rod guide tubeis shown in Fig. 2.18.

COUPLING Design Code: The control rod guide tube is designed toRELEASE meet the requirements of Sect. IlI of the ASME B&PVHANDLE Code.

Material or Construction: The control rod guide tube isVELOCITY made of type 304 stainless steel.LIMITER

COUPLINGSOCKET

ORNL.DWO 93-3406 ETM

Figure 2.17 Control blade

AngnmentK

Component No. 22: Vessel Head Cooling md

Spray Nozzle

Function: The vessel head cooling spray nozzle maintainssaturated conditions in the reactor head region by condens-ing steam being generated by the hot reactor vessel wallsand other internals. The spray also has the effect of reduc-ing thermal stratification in the reactor vessel coolant sothat the water level in the vessel can rise if needed. Ahigher water level can provide effective cooling to more ofthe reactor components inside the reactor vessel. . dr

Description: A part of the reactor coolant flow returning Contma od ddveto the reactor vessel can be diverted to a spray nozzle in the God St""

reactor head. The spray nozzle is mounted to a short lengthof pipe and a flange. The spray nozzle flange is bolted to amating flange on the reactor vessel head nozzle. Figure 2.18 Control rod guide tube

Design Code: The vessel head cooling spray nozzle isdesigned to meet the requirements of Sect. III of theASME B&PV Code.

NUREG/CR-5754 16

Component No. 24: Surveillance SampleHolders

Function: The surveillance sample holders contain impactand tensile specimen capsules. The test specimens are usedto measure irradiation effects on material properties of thereactor vessel wall materials.

Description: The surveillance sample holders are weldedbaskets hanging from brackets welded to the inside of thereactor vessel at the middle level of the core. The circum-ferential positions of the holders are chosen to expose thetest specimens to the same maximum neutron fluxes asthose experienced by the vessel wall.

Design Code: The surveillance sample holders are de-signed to meet the intent of Sect. III of the ASME B&PVCode.

Material of Construction: The surveillance sampleholders are made of type 304 stainless steel.

Component No. 25: Differential Pressure andLiquid Control Line

Function: The differential pressure and liquid control lineserves two functions: to sense the differential pressureacross the core plate and to provide a means for injectingliquid control solution into the coolant stream.

Description: The differential pressure and liquid controlline enters the reactor vessel as two concentric pipes at anelevation below the core shroud. The two pipes separate inthe lower plenum. The inner pipe ends near the lowershroud with a perforated section below the core plate. Theperforated pipe section senses the pressure below the coreplate during normal operations and also serves as the outletfor the injection of liquid control solution, as needed.Injection of the cooler liquid control solution through theinner pipe will reduce the effects of thermal shock on the

BWR

vessel nozzle. The outer pipe ends above the core plate andis used to sense the pressure in the region outside the fuelassemblies.

In some BWRs, the core spray line is used as the conduitfor injecting liquid control solution. In these reactors, thedifferential pressure sensing lines enter the reactor vesselthrough two bottom head penetrations. One line ends nearthe lower shroud with a perforated pipe section for sensingpressure below the core plate. The other line ends justabove the core plate and senses the pressure in the regionoutside the fuel assemblies.

Design Code: The differential pressure and liquid controlline is designed to meet requirements of Sect. Inl of theASME B&PV Code.

Material or Construction: The differential pressure andliquid control line is made of type 304 stainless steel.

References

1. American Society of Mechanical Engineers, "ASMEBoiler and Pressure Vessel Code, Section m, NuclearPower Plant Components, Div. 1," 1992.*

2. American Society of Mechanical Engineers, "ASMEBoiler and Pressure Vessel Code, Section IX, Weldingand Brazing:' 1992.*

3. Northern States Power Company and MultipleDynamics Corporation, "BWR Pilot Plant Life Exten-sion Study at the Monticello Plant Interim Phase 2,"Interim Report, EPRI NP-5836M, October 1988.

Available from American National Standards Institute, 1430 Broadway,New York, NY 10018, Copyrighted.

17 17 ~~~~~~NTUREG/CR-5754

3 Primary Stressors

Reactor internals are submerged in a hot and moving liquidcoolant during normal plant operations. The nominal sys-tem pressure inside the reactor vessel is 7.3 MPa(1055 psia). The coolant saturation temperature corre-sponding to the system pressure is 288*C (550'F). Theaverage core inlet flow velocity is about 2.1 m/s (7 ft/s).Internal components located in the vicinity of the core arealso exposed to fast neutron (E > I MeV) fluxes andgamma radiation. The operating environment inside a reac-tor pressure vessel generates many stressors that are con-sidered as necessary to the development of aging-related ortime-dependent degradation mechanisms.

The term "stressor" is used to represent conditions thatmay contribute to the initiation and the sustaining of thesubsequent growth of an aging-related degradation mecha-nism. Applied loading is probably the most commonstressor for structural components. In addition to appliedloadings, environmental effects and manufacturing pro-cesses can also impose stressors on reactor internals.Environmental effects and manufacturing processes maycause changes in the mechanical and physical properties ofthe materials of construction. Changes in material proper-ties can have a significant impact on the aging process.

Abnormal operating conditions can impose much moresevere conditions on reactor components. However, theywill not be investigated in the present study where theemphasis is on the assessment of aging effects under nor-mal plant operations. Normal plant operations constitutethe great majority of a reactor's operating history.

3.1 Applied Loadings

Applied loadings are mechanical and thermal loads associ-ated with normal and transient (startup or shutdown) plantoperations. External mechanical loads include static differ-ential pressure loadings, preloads in bolts, and hydro-dynamic forces produced by coolant flows in the pressurevessel. An important internally applied load is one causedby welding-induced residual stresses.

Flow-gencrated hydrodynamic forces can be static (steadystate) or oscillatory in nature. Steady-state hydrodynamicforces are usually referred to as lift and drag forces; theyare counterbalanced by the weight of the component andstructural supports. Oscillatory hydrodynamic forces aregenerated by flow separations. Pump-generated pressurepulsations can also act as periodic external excitations toreactor internals. In the BWR environment, oscillatoryhydrodynamic forces are the major concern because they

can cause a component to vibrate. FIV is a major cause offatigue problems in reactor internals.

Preloads on bolts are applied loads acting on reactor inter-nals. Tensile stresses produced by the preloads maybecome a contributing factor to the development of cracksin bolts. Most of the BWR internals are not part of the pri-mary pressure boundary and are not subjected to largestatic differential pressure loads. Exceptions are in-coreinstrumentation guide tubes and housings; differentialpressure loads are incorporated into design loadings forthese components.

Thermal loads are caused by the existence of temperaturegradients in a component, by thermal expansions of differ-ent materials, and by restricted thermal expansions. Reac-tor internals are designed to accommodate differentialthermal expansions; as a result, constraint-induced, steady-state thermal loads are kept at a low level. Some internalcomponents are exposed to the mixing of fluid flows at dif-ferent temperatures. The mixing actions produce rapidtemperature changes in the components. These thermalcyclings can lead to the development of alternating stressesand fatigue cracks.1

The applied loadings of major concern to BWR internalsare flow-generated cyclic loads. They produce alternatingstresses that can lead to fatigue failures.

3.2 Environmental Stressors

Major environmental stressors imposed on reactor internalsare the result of their contact with a hot and potentially cor-rosive coolant. The corrosiveness of the coolant is con-trolled primarily by the presence of dissolved oxygen inthe reactor cooling water. Dissolved oxygen is a product ofradiolytic reactions in the core. The level of dissolved oxy-gen in the BWR cooling water, 100 to 300 ppb, is suffi-cient to produce the electrochemical driving force neededto promote SCC in sensitized austenitic stainless steelcomponents. The electrochemical potential of an austeniticstainless steel is increased by the buildup of the dissolvedoxygen content in the cooling water, and the stainless steelbecomes more susceptible to corrosion. Type 304 stainlesssteel, the most common material of construction for inter-nals, is susceptible to SCC in the sensitized condition. Thepresence of other impurities such as chlorides may accel-erate the corrosion process.

The relatively high operating temperature can producephysical changes in some reactor internal components.

19 19 ~~~~~~NUREG/CR-5754

I

Primary

Parts made from cast austenitic stainless steel (CASS) canbecome embrittled after prolonged operations in the BWRoperating temperature.2 Embrittled materials are more sus-ceptible to fracture failures.

The exposure to fast neutron (E > 1 MeV) fluxes is anotherimportant environmental stressor for internal components.3

The effects are most noticeable for components located inclose proximity to the core. Long-term exposure to fastneutron fluxes can change the mechanical and physicalproperties of the materials of construction for internals.Irradiation effects can increase the yield and ultimatestrengths and decrease the ductility and fracture toughness,as well as the critical flaw size, of a material. Neutronirradiation effects can also lower the temperature at whichcreep can become a significant deformation mechanism.These changes may adversely affect the structural integrityof reactor internals.

3.3 Manufacturing Stressors

Methods for fabricating reactor internals may impose stres-sors on the components. Manufacturing stressors can alsopromote the development of aging-related degradationmechanisms. Welding and cold-working are two commonmanufacturing methods used in making internals, and theycan introduce stressors to the finished parts.

The conventional welding process sensitizes austeniticstainless steels (such as type 304) and makes them suscep-tible to corrosion. Residual stresses in weld heat-affectedzones (HAZs) may contribute to the development of SCC.

Chromium and nickel are the two major alloying elementsused in the making of austenitic stainless steels. Carbon isalso used as a minor alloying element. Austenitic stainlesssteels, such as type 304, have generally good corrosionresistance. However, sensitization can make a weldedaustenitic stainless steel component vulnerable to SCC.When a structural component made of an austenitic stain-less steel is cooled down slowly through the temperaturerange from -820 to 4800C (1500 to 9001F), as in a weldingprocess, chromium carbides precipitate at grain boundaries.The precipitation of chromium carbides will lead to adecrease in the chromium content in regions adjacent tograin boundaries. When the chromium content in type 304stainless steel drops below 12%, the steel becomes suscep-tible to corrosion attacks at the grain boundaries.4 Thechromium depletion process is known as the sensitizationprocess, and weld HAZs are common locations for sensi-tized austenitic stainless steel.

In addition to the material sensitization process, residualstresses are also generated in weld HAZs during cooldown.Residual stresses may contribute to the development ofSCC.

Cold-working introduces new manufacturing stressors.Plastic strains are accumulated in the workpiece duringcold-working, and excessive plastic strain accumulationcan lead to the development of cracks. Contacts with a hotand corrosive coolant may accelerate the crack initiationprocess. Evidence also suggests that cold-working caninduce martensite formation in the workpiece. Martensiteformation may help the material sensitization process andmake a cold-worked component more susceptible to corro-sion attacks.5

Applied loadings, environmental effects, and manufactur-ing processes create stressors that, by themselves or in con-junction with each other, create conditions favorable to thedevelopment of aging-related degradation mechanisms inBWR internals.

References

1. E. Kiss and T. L. Gerber, "Status of Boiling WaterResearch Structural Integrity Program at the GeneralElectric Company," Nucl. Eng. & Des. 59,27-45(1980).*

2. 0. K. Chopra and H. M. Chung, "Long-Term Aging ofCast Stainless Steel: Mechanisms and ResultingProperties," Trans. Fifteenth lWater Reactor SafetyResearch Information Meeting, Gaithersburg, Md.,USNRC Conference Proceedings, NUREG/CP-0090,October 1987.t

3. R. E. Robins, J. J. Holmes, and J. E. Irvin, "Post Irra-diation Tensile Properties of Annealed and Cold-Worked AISI-304 Stainless Steel," Trans. Amer. NucL.Soc. (November 1967).*

4. R. L. Cowan and C. S. Tedmon, "Intergranular Cor-rosion of Iron-Nickel-Chromium Alloys," Adv.Corros. Sci. Tech. 3 (1973).*

5. "BWR Pilot Plant Life Extension Study at the Monti-cello Plant Interim Phase 2," Interim Report, EPRINP-5836M, Northern States Power Company andMultiple Dynamics Corporation, October 1988.

Available in public technical libraries.tAvailable for purchase from National Technical Information Service,

Springfield, VA 22161.

20NUREG/CR-5754

4 Aging-Related Degradation Mechanisms

The presence of stressors is essential to the development ofaging-related degradation mechanisms. Major potentialaging-related degradation mechanisms for BWR internalsare identified as SCC, fatigue, embrittlement, erosion,creep, and stress relaxation. SCC includes intergranularSCC (IGSCC) and IASCC. Fatigue is caused by vibrationsand thermal cycling. Embrittlement can develop as a resultof prolonged exposure to fast neutron fluxes and thermalaging. Contacts with a high-speed liquid flow or flows withdroplet impingements are common causes of erosion prob-lems. Creep and stress relaxation are caused by prolongedexposure to high temperatures and fast neutron fluxes.They can lead to changes in material properties and defor-mation mechanisms. Stressors associated with these aging-related degradation mechanisms are present inside a BWRpressure vessel.

4.1 Stress Corrosion Cracking

Corrosion is the weakening of a structural component as aresult of material deterioration caused by electrochemicalreactions with the surrounding medium. The effects can beglobal or highly localized. Global effects are referred to asgeneral corrosion. The localized effects usually involvesome form of crack development.

The surface of a metallic structure can become oxidizedwhen it is submerged in a corrosive fluid medium. The re-moval of corrosion products by the fluid medium will re-sult in a thinning of the wall. This general corrosion pro-cess is global and occurs over the entire contact surface ina somewhat uniform manner. Austenitic stainless steels,such as type 304, have good resistance to general corrosionand wastage.

SCC is a common form of highly localized corrosion phe-nomena. SCC can occur in ductile materials with little orno plastic strain accumulation associated with the process.The development of SCC in a structural component re-quires the simultaneous presence of three conditions: aconducive environment, a susceptible material, and tensilestresses above a threshold level. SCC is not likely to de-velop when any one of these three conditions is absentfrom the operating environment. Thus, the elimination ofone condition is the basis for formulating strategies to con-trol SCC. Depending on the alloy compositions and the na-ture of stressors presented in the system, cracks can de-velop along grain boundaries; such SCC is known asIGSCC. When cracks propagate along certain crystallo-graphic slip planes within the grains, the failures areknown as transgranular stress corrosion cracking

(TGSCC). More detailed discussions on the fundamentalsof SCC can be found in texts by Logan1 and Romanov.2

The hot and oxygenated reactor cooling water creates acorrosive environment in the BWR pressure vessel. Thedissolved oxygen in the reactor cooling water increases theelectrochemical potential of type 304 stainless steels andmakes them vulnerable to corrosion attacks. The presenceof impurities, such as chlorides and sulfates, in the coolantmay accelerate the crack development process.

Most of the SCC failures in BWR internals are found inweld HAZ. In addition to the oxygenated reactor coolingwater, the welding process can provide the other two con-ditions that are needed for the development of SCC.

When a weld is cooled down through the temperaturerange from 820 to 4800C (1500 to 9000F), type 304 stain-less steel undergoes a sensitization process characterizedby a chromium depletion at grain boundaries.3 The sensi-tization process makes austenitic stainless steels suscepti-ble to corrosion attacks. The presence of residual stressesin weld HAZs supplies the third requirement for SCC. Thelevel of residual stresses may be reduced by proper heattreatment Proper heat treatment for internals is not alwayspossible because of their large size. As a result, residualstresses remain in weld HAZs. Residual stresses are self-equilibrating, and tensile stresses are always present inthem. The threshold tensile stress level that is needed forthe development of SCC varies, but it is generally acceptedthat the level is near the yield stress of the material.Residual stresses in weld HAZs can approach the yieldstress value, and they can contribute to the development ofSCC. Tensile stresses generated by preloads in bolts canalso attain the threshold value needed for the developmentof SCC.

The increase in electrochemical potential of the material,the chromium depletion sensitization process, and the pres-ence of tensile stresses provide the three necessary condi-tions needed for the development of SCCs. SCC is proba-bly the most common aging degradation mechanism inBWR internals. The majority of SCC failures observed inBWR internals are intergranular. Other conditions mayalso have some influence on the development of SCC. Themore significant ones are crevice condition and prolongedexposure to fast neutron fluxes.

A crevice is a general term describing a small and narrowregion containing a stagnant fluid. High concentrations of

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impurities (such as chloride and sulfate) may be trapped inthe stagnant fluid, creating a local environment that isfavorable to the development of SCC. The impurities pro-duce a highly acidic solution that is anionic in nature andcan destroy the protective thin oxide film and increase thelikelihood of corrosion attacks. The presence of crevices inan internal component can be a design feature, or they canactually be small cracks initiated by other degradationmechanisms. Narrow gaps between contact surfaces, suchas those found in bolted joints, are design features that maycreate crevices in an internal component.

SCC has been observed in nonsensitized stainless steel in-ternal components. In some cases, the stress levels in thesecomponents are relatively low and below the material yieldstress. A common factor in these nonsensitized, low-stressinternal components with SCC failures is a prolonged ex-posure to fast neutron (E > I MeV) fluxes. Reactor inter-nals that are exposed to fast neutron fluxes in a corrosiveenvironment are susceptible to IASCC.4 Most IASCCs areintergranular.

Basic understanding of IASCC is not complete, and nounified theory can account for all radiation effects on SCC.Evidence suggests that a threshold neutron fluence levelexists below which IASCC is not likely to occur. The bestestimate for the threshold value is about 5 x 1020(neutrons/cm2) (Ref. 5). BWR internals, such as controlblades, OFS pieces, the core shroud, the core plate, the topguide, and start-up neutron sources can experience lifetimeneutron fluence levels exceeding the threshold values.These components are susceptible to IASCC.

It has been suggested that irradiated materials are weak-ened by gas bubbles formed by transmutation processes.6

Trace quantities of boron, which can react with thermalneutrons to form lithium and helium,6 are found inaustenitic stainless steels. Hydrogen is also produced by atransmutation process involving fast neutrons and elementssuch as N, Ni, and Cr (Ref. 4). Bubbles are formed whenthese gases precipitate from the solid. The bubbles have atendency to move to dislocations and grain boundaries,6

and they create discontinuities that can weaken the mate-rial. The solid may disintegrate in the absence of hightensile stresses when the bubbles are fused together. Whilethe gas bubble formation mechanism can account for someof the observed characteristics of IASCC, questions remainon bubble sizes, their weakening effects on the solid, andthe driving forces behind the bubble movement

The presence of impurities such as phosphorous andsilicone increases the susceptibility of unirradiatedstainless steels to IGS CC. It has been suggested that high-

NUREG/CR-5754

purity stainless steels are more resistant to IASCC thancommercial-grade stainless steels.4 However, conclusiveevidence to support this assertion is lacking.

The basic understanding of IASCC is not complete, andthere is still much active research work in the field.General agreement is that the exposure to fast neutron flu-ence above a threshold value, together with a corrosiveenvironment, can provide conditions that are favorable tothe development of IASCC. The threshold tensile stresslevel needed for the development of IASCC may besignificantly lower than the material's yield stress.In summary, most welded BWR internals are susceptible toSCC. Crevice conditions may aid the SCC process.Prolonged exposure to fast neutron fluxes can lead toIASCC in nonsensitized steel components with low stresslevels. Of the 25 selected BWR internals, 18 operate inconditions that would make them susceptible to SSC; theyare summarized in Table 4.1. The last six componentslisted in Table 4.1 are also vulnerable to IASCC.

4.2 Fatigue

Reactor internals are subjected to time-dependent ordynamic loads. A structure will vibrate as a response todynamic loads. Vibrations can lead to crack developments,and structural failures caused by vibrations are known asfatigue failures. Cyclic loadings and flow-induced vibra-tions are the root causes of fatigue problems in BWR inter-nals. The loadings can be either mechanical or thermal innature. Vortex shedding and other unsteady flow effectscan generate oscillatory hydrodynamic forces, while cyclicthermal loads are produced by contact with the turbulentmixing of flow streams at different temperatures. Fatiguefailures can be divided into two types: high- and low-cyclefatigues. When stress is in the linear-elastic range and theamplitude of the vibration is small, the resulting failure isclassified as high-cycle fatigue. Little or no plastic strainaccumulation is associated with high-cycle fatigue failures.High stress values, large amplitude vibrations, and signifi-cant plastic strain accumulation are important characteris-tics associated with low-cycle fatigues. In addition to thestress and strain criteria, the number of cycles of vibrationsbefore a failure occurs is used to identify low- and high-cycle fatigue. When a crack is initiated between 103 and104 cycles, the failure is classified as a low-cycle fatiguefailure.

The fluid flows inside the reactor pressure vessel can im-pose mechanical and thermal loads to internals submergedin the flow streams. A common mechanical load is the dragforce, which is a steady-state hydrodynamic force acting ona submerged structure. In transient operations, the propa-gation of pressure waves through the system can generate

;

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Table 4.1 BWR Internals susceptible to SCC

Component Stressors

Steam dryer supportring Corrosive coolant, cold working, weld HAZShroud head bolts - Corrosive coolant, bolt preloads, crevice conditionsAccess hole cover Corrosive coolant, crevice conditions, weld HAZSteam separator Corrosive coolant, weld HAZCore spray sparger - Corrosive coolant, weld HAZ, cold-workingShroud head Corrosive coolant, weld HAZCRD housing Corrosive coolant, weld HAZFeedwater sparger Corrosive coolant, weld HAZCore spary line internal piping Corrosive coolant, weld HAZ, crevice conditionsJet pump Corrosive coolant, preloads, weld HAZIn-core neutron flux monitor housing Corrosive coolant, weld HAZIn-core neutron flux monitor guide tubes Corrosive coolant, weld HAZIn-core neutron flux monitor dry tubes Corrosive coolant, weld HAZ, fast neutron fluxesNeutron source holder Corrosive coolant, fast neutron fluxes, crevice con-

ditionsCore shroud Corrosive coolant, fast neutron fluxes, weld HAZCore plate Corrosive coolant, weld HAZ, fast neutron fluxes,

crevice conditionsControl blades Corrosive coolant, fast neutron fluxes, weld IIAZ,

- crevice conditionsTop guide Corrosive coolant, fast neutron fluxes, crevice con-

ditions,OFS pieces Corrosive coolant, fast neutron fluxes

oscillatory hydrodynamic forces. Core internals are de-signed to withstand such steady-state and transient loads.The primary cause of high-cycle fatigue failures in BWRinternals is attributed to a phenomenon known as FIV.When a fluid medium flows over a blunt body, flow sepa-rations lead to the shedding of vortices from the body.Vortex shedding generates a cyclic load acting on thestructure that would cause it to vibrate. In most cases, theamplitudes of the vibrations are small. A structure that issubjected to small-amplitude vibrations is susceptible tohigh-cycle fatigue failures. In rare instances, destructivelarge-amplitude resonant vibrations could develop in thestructure when its fundamental natural frequency coincideswith an input excitation frequency. However, reactor inter-nals are designed to avoid operating under a resonant.vibration condition. Key reactor internal components aremonitored during reactor preoperation testings to ensurethe absence of flow-induced resonant vibrations.

The operation of pumps in the reactor primary coolant sys-tem introduces flow-related excitations into the reactorcoolant flow stream. They are pressure pulsations withdominant frequencies closely related to the pump rotating

speed, blade passing frequency, and their harmonics. Theeffects of these pressure pulsations are strongest at the en-trance regions around vessel inlet nozzles. Coincidence ofa dominant pressure pulsation frequency with a structuralfundamental frequency can also lead to resonant vibrationproblems. In any event, these pump-generated pressurepulsations are excitation sources for small-amplitude vibra-tions for reactor internals.

There are regions inside the reactor pressure vessel wherereactor coolant flow streams at different temperatures aremixed in turbulent conditions. Reactor internal componentslocated in these regions are exposed to a fluid medium withrapidly changing temperatures. The changing temperaturescan impose thermal loads on components submerged in themixing flow streams. The effects of the varying stressescaused by thermal cycling loads are similar to those causedby vibrations. The components are susceptible to high-cycle fatigue failures.7 Usually little or no plastic strainaccumulation occurs in a structure undergoing rapidthermal cyclings.

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Low-cycle fatigue failures are mostly associated with ex-cessive structural deformations under cyclic applied loads.The components are usually undersized for the loads thatthey are required to carry, and plastic strains are accumu-lated as a result of large structural deformations. Failure,generally in the form of crack development, will occur ifthe plastic strain accumulations exceed a critical value.Strengthening or "beefing up" the component is a commonand effective solution to low-cycle fatigue problems.

BWR internals that are subjected to cyclic loadings aresusceptible to fatigue failures. Major cyclic loadings in-clude flow-generated oscillatory hydrodynamic forces andthermal cyclings caused by turbulent mixing of coolantflow streams at different temperatures. Of the 21 reactor in-ternals, 6 are susceptible to fatigue failures and are summa-rized in Table 4.2.

Table 4.2 BWR internals susceptible to fatigue

Component Stressors

Jet pump sensing line Flow-generated oscillatory,hydrodynamic forces

Steam dryer Large applied cyclic pressureloads

Feedwater sparger Flow-generated oscillatoryhydrodynamic forces, leakageflow-induced rapid thermalcyclings

Core spray sparger Flow-generated oscillatoryhydrodynamic forces

Jet pump Flow-generated oscillatoryhydrodynamic forces

In-core neutron flux Flow-generated oscillatorymonitor housings hydrodynamic forces

In-core neutron flux Flow-generated oscillatorymonitor guide tubes hydrodynamic forces

In-core neutron flux Flow-generated oscillatorymonitor dry tubes hydrodynamic forces

4.3.1 Radiation Embrittlement

In addition to fast neutron fluence level, effects of radiationembrittlement are also determined by the irradiation tem-perature and material compositions. Austenitic stainlesssteels, such as type 304, are considered susceptible to ra-diation embrittlement. The assessment of radiation embrit-tlement effects is based primarily on the interpretations ofexperimental results. The experimental data base is ob-tained from testings using surveillance specimens fromcommercial power reactors and other test reactor experi-ments. The reduction in uniform elongation is used as anindication of the decrease of ductility in a material after aprolonged exposure to fast neutron fluxes. Changes in frac-ture toughness are deduced from results of Charpy V-notchimpact testings and other types of fracture-toughness test-ings.

Experiments8 were conducted at a temperature range fromroom temperature to -7600 C (1400'F) and fast neutron(E > 1 MeV) fluence levels up to about 6 x 1021 neutrons/cm2. At -3000C (5700F), the unirradiated uniform elonga-tion for type 304 stainless steel is -38%. At the same testtemperature, the uniform elongation was reduced to -22%at a neutron fluence of -1.5 x 1020 neutrons/cm2. At -1.5x 1021 neutrons/cm2 the uniform elongation at 300'C wasfurther reduced to -0.5%. The decrease in uniformelongation seemed to level off at higher neutron fluencelevels. At a temperature of -300'C (570'F), testing results 8

using type 304 stainless steel test specimens indicated thateffects of radiation embrittlement begin to manifest at aneutron fluence level of -5 x 1020 neutrons/cm2. Theexpected fast neutron fluence for reactor internals isdetermined by the distance of the components from thecore. In 40 years of power operations, the estimated maxi-mum neutron fluence is -1 x 1022 neutrons/cm 2 and isfound at the central region of the top guide.9 The esti-mated 40-year neutron fluence levels for other componentslocated close to the core,9 such as the core plate, coreshroud, and fuel support pieces, range from -3 x 1020 to5 x 1021 neutrons/cm2. These internal components aresusceptible to effects of radiation embrittlement. Theexpected 40-year fast neutron fluence levels for otherinternals are below 1.5 x 1020 neutrons/cm 2, and theirmaterials of construction should retain sufficient ductilityand fracture toughness to ensure that brittle fracture is notlikely to occur.

4.3 Embrittlement

Embrittlement is the loss of ductility and fracture tough-ness of a material, and it is usually accompanied by in-creases in the material yield and ultimate strengths.Prolonged exposures to fast neutron fluxes and high tem-peratures (thermal aging) can cause embrittlement in com-ponents made of stainless steels. The process can transforma ductile material to one that is susceptible to brittle frac-ture failures.

NUREG/CR-5754 224

4.3.2 Thermal Embrittlement

Thermal embrittlement is caused by a prolonged exposureto high temperatures. It is a time-dependent degradationmechanism that affects mostly reactor internals made ofCASS.10 CASS is a two-phase alloy composed of austeniteand ferrite. The determining factor is the ferrite content inthe alloy; when it is <20%, CASS is not sensitive to ther-mal embrittlement. When the ferrite content is >20%, andwhen the operating temperature is in the range from 427 to4821C (800 to 9000F), a chromium-rich precipitation in theferrite phase would lead to a decrease in the fracturetoughness of the material. Also indications are that the fer-rite phase could become brittle when a CASS component isaged at a temperature of -3161C (600oF). Thermal aging isa potential aging-related degradation mechanism becausethe ferrite content in CASS BWR internal components cango up to 25%, and the required temperature ranges arewithin the service conditions inside the pressure vessel.BWR internal components that are susceptible to radiationand thermal embrittlements are summarized in Table 4.3.

Table 4.3 BWR Internals susceptible toembrittlement

Agingtwo potential problem areas associated with erosion inBWR internals are bubble formation in a liquid flowstream and liquid droplet impingements in two-phaseflows.

Bubbles in a liquid flow stream can become attached tosolid surfaces and can cause erosion by cavitation. In addi-tion, the formation and collapse of bubbles can create dis-turbances in the flow that can disrupt the protective bound-ary layer film adjacent to solid surfaces and expose freshmetal surfaces to the erosion process. High-speed flow re-gions, such as those in the jet pump throat, nozzle, anddiffuser, are vulnerable to bubble-induced erosion.

Erosion also can occur in a two-phase flow regime, such asthe steam-water flows in steam separators. The impact ofliquid droplets on a metal surface is analogous to theimpact of solid particles and can cause abrasion problems.Of the 25 selected BWR internals, 2 are susceptible toerosion and are summarized in Table 4.4.

Table 4.4 BWR internals susceptible to erosion

Component Stressors

Top guide Fast neutron fluxesCore plate Fast neutron fluxes

Core shroud Fast neutron fluxesControl blades Fast neutron fluxesIn-core neutron flux Fast neutron fluxes

monitor dry tubesCast components of the Thermal aging and ferrite

jet pump assembly precipitation in CASSOFS pieces Fast neutron fluxes, ther-

mal aging, and ferriteprecipitation in CASScomponents

Cast components of the Thermal aging and ferritesteam separator as- precipitation in CASSsembly

Component Stressors

Throat, nozzle, and dif- Cavitation-inducedfuser sections in jet, bubblespump

Steam separator Liquid dropletimpingements

4.5 Creep and Stress Relaxation

Creep is the progressive deformation of a structure under aconstant stress. Stress relaxation is the reduction of internalstresses in a component under constraints. Creep can leadto fracture failures in a structural component. Stress relax-ation can loosen bolts in bolted joints, which, in turn, couldresult in a loss of structural integrity and leakage problems.

Creep and stress relaxation are generally associated withstructures that operate in a high-temperature environment.Exposures to fast neutron fluxes may also affect the devel-opment of creep and stress relaxation: As a general rule,when the operating temperature is less than half of themelting point temperature of the material, effects of creepand stress relaxation are not significant. The melting pointtemperature for a typical austenitic stainless steel is-1430'C (2600'F). The normal operating temperature forBWR internals is about 2880C (550'F), less than half ofthe stainless steel melting point temperature. Thermally

NUREGICR-5754

4.4 Erosion

Erosion is a potential aging-related degradation mechanismwhen an internal component comes into contact with liquidflows. The abrasive effects are well-known when solid par-ticles are present in a flow stream. Reactor cooling water isof high purity; solid particles, if they are present in theflow stream, will be very small in quantity. Solid particleabrasion is not a serious problem for BWR internals. The

25

Aging

induced creep and stress relaxation are not considered asmajor aging-related degradation mechanismns for reactorinternals.

Prolonged exposure to fast neutron fluxes lower the tem-perature at which creep and stress relaxation can becomesignificant deformation mechanisms. Results of in-pile ex-periments conducted at 2880C (5500F) showed that effectsof stress relaxation can be observed in bolts made of type304 stainless steelsl at a neutron fluence of 6 x 1019 neu-trons/cm 2. At about the same temperature, significantstress relaxation has been observed in type 304 stainlesssteel test specimens at neutron fluence above 5 x 1020 neu-trons/cm 2. BWR internals located in the vicinity of thecore have 40-year neutron fluence levels in the range from6 x 1019 to 5 x 1020 neutrons/cm2. Of the 25 selectedcomponents, 4 are susceptible to creep and stress relax-ation and are summarized in Table 4.5.

Table 4.5 BWR internals susceptible to radiation-Induced creep and stress relaxation

Component Stressors

Top guide Fast neutron fluxes and hightemperature (2880C)

Core shroud Fast neutron fluxes and hightemperature (2880C)

OFS pieces Fast neutron fluxes and hightemperature (2880C)

Control blades Fast neutron fluxes and hightemperature (2880C)

In-core neutron flux Fast neutron fluxes and highmonitor temperature (2880C)dry tubes

2. V. V. Romanov, "Stress Corrosion Cracking ofMetals," The Israel Program for Scientific Translationand the National Science Foundation, Washington,D. C., 1961.

3. R. L. Cowan and C. S. Tedmon, "IntergranularCorrosion of Iron-Nickel-Chromium Alloys," Adv. inCorr. Sci. & Tech 3 (1973).

4. A. J. Jacobs, and G. P. Wozaldo, "Irradiation AssistedStress Corrosion Cracking as a Factor in NuclearPower Plant Aging," J. Mat. Eng. 9,4 (1988).*

5. Northern States Power Co. and Multiple DynamicsCorp., "BWR Pilot Plant Life Extension Study at theMonticello Plant: Interim Phase 2," EPRI NP-5836M,October 1988.

6. J. Gittus, Irradiation Effects in Crystalline Solids,Applied Science Publisher, London, 1978.

7. E. Kiss and T. L. Gerber, "Status of Boiling WaterReactor Structural Integrity Program at the GeneralElectric Company," Nucl. Eng. Des. 59, 27-45(1980).*

8. R. E. Robins, J. J. Holmes, and J. E. Irvin, "PostIrradiation Tensile Properties of Annealed and Cold-Worked AISI-304 Stainless Steel," Trans. Amer. Nucl.Soc. (November 1967).

9. V. S. Shah and P. E. MacDonald, Eds., "Residual LifeAssessment of Major Light Water ReactorsComponents-Overview," USNRC ReportNUREG/CR4731, November 1989 .t

10. 0. K. Chopra and H. M. Chung, "Long-Term Agingof Cast Stainless Steel: Mechanisms and ResultingProperties," Trans. Fifteenth Water Reactor SafetyResearch Information Meeting, Gaithersburg, Md.,USNRC Proceeding NUREG/CP-0090, October1987.t

11. Structural Integrity Associates, Inc., "Component LifeEstimation: LWR Structural Materials DegradationMechanisms," Interim Report, EPRI NP-5461,September 1987.

In summary, 19 of the 25 selected BWR internals are sus-ceptible to SCC, including IASCC. Specifically, 6 out ofthe 19 are susceptible to IASCC. Of the 25 components, 7are vulnerable to fatigue failures. Embrittlement is a poten-tial aging-related degradation mechanism for eight compo-nents, and erosion is a potential aging-related degradationmechanism for two internal components. Five componentsare susceptible to the effects of creep and stress relaxation.The 25 selected components and their potential aging-related degradation mechanisms are shown in Table 4.6.

References

1. H. L. Logan, The Stress Corrosion of Metals, JohnWiley & Sons, New York, 1966.

Available in public technical libraries.tAvailable for purchase from National Technical Information Service,

Springfield, VA 23161.

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Table 4.6 BWR internals and potential aging-related degradation mechanisms

Component SCC Creep Fatigue Embrittlement Erosion

Steam dryer .. _____ *

Steam separator * *

Shroud head *

Shroud head bolts *

Steam separator support ringTop guide * *

Access hole cover *

Core shroud * *

OFS piece *

Core plate

Core spray line internalpiping

Core spray sparger

Feedwater sparger* *

Jet pump * * * *

In-core neutron flux monitorhousings

In-core neutron flux monitor * *

guide tubes .

In-core neutron flux monitor * * *

dry tubesCRD housingNeutron source holder _Jet pump sensing line *

Control blade

27 27 ~~~~~~~NUREG/CR-5754

5 Survey of Aging-Related Failures

Stressors and potential aging-related degradationmechanisms have been identified for the 25 selected BWRinternals. Aging-related degradation mechanisms do notdevelop at the same rate in the environment inside thereactor vessel. Therefore, component failure information isneeded to properly assess the relative importance of the po-tential aging-related degradation mechanisms. The identifi-cation of major aging-related degradation mechanisms canalso provide information that can be used to formulatestrategies for managing aging effects.

As components of an operating nuclear power reactor, re-actor internals are regularly inspected as part of the plantISI program. The ISI programs for domestic nuclear powerplants are established under the rules and regulations ofSect. XI of the ASME B&PV Code.1 The Code is also thebasis for ISI programs in most countries that utilize U.S.-developed reactor technologies.2

The program outlines specific procedures and requirementsfor inspecting reactor components, structures, and systemsto ensure safe plant operation. The programs may includeorders, rules, criteria, and guidelines established by regula-tory agencies as well as industry groups.

5.1 ISI Program for Reactor Internals

The plant ISI program calls for the visual inspection of re-actor internals. A complete inspection cycle is 10 years,and selected components are inspected at refueling out-ages.

Section XI of the ASME B&PV Codel specifies threeclasses of visual inspections: VT-1, VT-2, and VT-3. AVT-I examination is conducted to determine the conditionof the part, component, or surface examined, includingsuch conditions as cracks, wear, corrosion, erosion, orphysical damage on the surface of the part or component.The examination can be performed either directly or re-motely.

A VT-2 examination is used to detect leakage from pres-sure retaining components. Most reactor internals are notpressure boundary components, and, as a result, VT-2 in-spections are seldom used on internals.

A VT-3 examination is conducted to determine the generalmechanical and structural conditions of components andtheir supports, such as verifications of clearances, settings,

physical displacements, loose or missing parts, debris,corrosion, wear, erosion, or the loss of integrity at bolted orwelded connections. VT-3 inspections can be performedeither directly or remotely. More detailed information ofthe inspection methods and procedures can be found in theappropriate subsections of Section XI of the ASME B&PVCode.1

During a refueling outage, some internal components areremoved from the pressure vessel and stored in a pool.VT-i and VT-3 visual examinations are performed on ac-cessible areas of reactor internals when they are in the stor-age pool. The underwater inspection is performed with re-mote television cameras under proper lighting. Internalsthat remain in the vessel are also visually examined by re-mote television cameras.

When a crack or flaw is detected, detailed informationconcerning the crack, such as its location and suspectedcauses, are recorded. The information is also sent to NRCas a Licensee Event Report (LER). Information on correc-tive actions taken are also included in the LER. For thesereasons, LERs are considered reliable sources for reactorcomponent failure information.

5.2 Reported Failure InformationSurvey

The survey of the aging-related component failure informa-tion for BWR internals is compiled from LERs and non-proprietary EPRI reports on nuclear power reactor operat-ing experiences. Most of these failures were reported in a10-year period from 1980 to 1989. It is difficult to extractreactor internal failure information from LERs, and it ispossible that a few failure cases may have been missed.Therefore, the survey results are not exhaustive in nature,but they should provide a representative picture of theaging-related failure situation for BWR internals.

Because of access limitation problems, four BWR internalsare not inspected: the OFS piece, steam separators, shroudhead, and core plate. The other 21 components were in-spected in accordance with the plant ISI program. Aging-related failures were detected in 19 of the 21 components.The survey results indicated that the core shroud and thetop guide are the two regularly inspected components thathad no reported failures during the survey period.However, the core shroud and the top guide are subjectedto many significant stressors and are considered vulnerableto aging-related degradations.

29 29 ~~~~~~~NUREG/CR-5754

Survey

5.3 Internal Component FailureInformation Survey Results

A review of the survey results indicates that most reportedaging-related failures in the 10-year survey period can beattributed to two aging-related degradation mechanisms:SCC (including IASCC) and fatigue. Radiation-inducedembrittlemcnt may have contributed to the failures of com-ponents that were exposed to high-energy neutron fluxes.

5.3.1 Reported SCC Failures

Of the 25 selected BWR internals, 19 are susceptible toSCC. SCC was detected in 11 of the 19 components. Mostof the cracks were located at weld HAZs. The followingsections provide a brief description of the more significantreported failures.

53.1.1 Jet Pump Holddown Beanis

SCC was detected in jet pump holddown beams in sixBWRs in the early 1980s. The beam is made of a nickel al-loy (Inconel X-750). The cracks were located across theligament of the beam section at the mean diameter of thebolt thread area as illustrated in Fig. 5.1. The cracks wereintergranular. Low heat-treatment temperature (8850C),which led to the sensitization of the beam material, andhigh pre!oads on the beam bolt were major contributingfactors to the development of SCC in jet pump holddownbeams. Replacement holddown beams and beams in latermodels of BWRs are solution heat-treated at a higher tem-perature (1090'C), and preloads in the beam bolts werealso reduced. These measures were effective in preventingIGSCC problems in new holddown beams.-

A failure of the holddown beam could result in the disas-sembly of the jet pump and may also lead to a reduction of

ORNL-DWG 92-2913 ETM

the operation safety margin during postulated accident sit-uations. NRC has issued IE Bulletin No. 80-07 (Ref. 3) onIGSCC problems in jet pump holddown beams; also, thebulletin established requirements for the inspection of jetpump holddown beams during refueling outages.

5.3.1.2 Core Spray Spargers

Core spray sparger crackings were detected in six BWRs.The cracks were initiated and propagated by IGSCC. Coldworking and sensitization during the fabrication of thespargers and stresses incurred during installation wereconsidered as major factors in the development of the ob-served cracks.

Cracks in the core spray spargers could alter the flow dis-tribution to the core and might lead to the generation ofloose parts in the pressure vessel. NRC IE Bulletin No. 80-13 (Ref. 4) addresses issues concerning the core spraysparger cracking problems and outlines requirements forthe inspection of core spray spargers during refueling out-ages.

5.3.1.3 Access Hole Covers in Shroud Support

Ultrasonic (UT) inspections of one BWR showed indica-tions of partial through-the-wall cracks in the welds attach-ing the access hold covers to the shroud support plate.These cracks were not detected by visual inspections.Welding-induced residual stresses and crevice conditionson the weld are major contributing factors to the initiationand propagation of cracks in the access hole cover toshroud support welds.

A failure of the access hole cover to shroud support weldcould lead to a separation of the cover plate from theshroud support. The potential exists that a severed accesshole cover may be swept into the recirculation pump suc-tion line, causing damage to the pump. The unplugging ofthe access hole covers will also have the undesirable ef-fects of creating an alternate flowpath that would allowsome of the recirculating system flow to bypass the coreduring normal and accident operating conditions. NRC hasissued Information Notice No. 88-03 (Ref. 5) on the impor-tance of inspecting the access hole cover to shroud supportwelds, but the Information Notice did not specify any spe-cific inspection requirement.

53.1.4 In-Core Neutron Flux Monitor Dry Tubes

Indications of cracks were detected by visual inspections ofIRM and SRM dry tubes in several BWRs. The cracks

TRUNNION

Figure 5.1 Crack locations of JET pump holddownbeams

NUREG/CR-5754 330

were located in the thin tube segment surrounding the -compression spring. Due to the high neutron fluence levelat the locations of these components, it is generally ac-cepted that IASCC and radiation-induced embrittlementmay have contributed to the development of the observedcracks. The accumulation of cruds and oxide formation increvices between the guide plug and the thin tube segmentmay have produced the tensile stresses needed for the de-velopment of SCC.

The thin tube is a nonpressure-retaining boundary of thedry tube. There was no indication of cracks in the adapter,shaft, guide plug, and thick instrument cavity (a primarypressure retaining boundary). Cracked dry tubes are re-placed during refueling outages.

53.1.5 Stub Tube to CRD Housing Welds

During routine inspections of the pressure vessel of twoolder BWRs, leakage was detected in the gap between thevessel wall and the CRD housing. Reactor cooling water isleaked to the outside of the pressure vessel through a slip-fit clearance between the inside surface of the stub tubeand the outside surface of the CRD housing. The leak isconsidered as unisolable.

Leakage was attributed to the development of through-the-wall cracks in the stub tube in the vicinity of the J-weldthat joins the CRD housing to the top of the stub tube. Thecracks were intergranular. A typical crack is illustrated inFig. 5.2.


Survey

The leakage was stopped by using mechanical seals sizedto slip over the outer surface of the stub tube and by roll-swaging the CRD housing to close the gap with the vesselwall.

Stub tubes are parts of the reactor pressure vessel, and theirfailures are included in the report because they were causedby the attachment of the CRD housing. This is consideredas a generic problem for older BWRs with stub tubes. Stubtubes are not used in newer BWRs where the CRD housingis welded directly to the bottom bead of the reactor vessel.A search of the reported failure data base did not locate anyfailure of the CRD housing to vessel bottom head welds.

Other SCC failures were also reported and are summarizedin Table 5.1. The type of reactors involved and the numberof failure cases were not available when the informationwas taken from EPRI reports.

Specifically, IASCC failures were reported in three BWRinternals: control blades, IRM and SRM dry tubes, andneutron source holders. Radiation-induced embrittlementmay also have contributed to the failures of the dry tubes.

5.3.2 Reported Fatigue Failures

Of the 25 selected BWR internals 7 are susceptible to fa-tigue, and failures were detected in 6 of the 7 components.Most of the reported cases are classified as high-cyclefatigue caused by flow-induced vibrations. Low-cycle fa-tigue was the cause of failure of cracks detected in thinsteam dryer hoods, and excessive structural deformationsunder cyclic pressure loads were the responsible failuremechanism. The following is a brief description of themore important fatigue failure cases.

53.2.1 Feedwater Sparger

In 1972, large circumferential cracks were detected in afeedwater sparger segment in a BWR-3 unit. Subsequentinspections revealed similar cracks in other BWR-3 and -4units. The cracks were located near the sparger water inletin the vicinity of the feedwater nozzle. Failures were at-tributed to flow-induced vibrations and rapid thermal cy-cling.

The thermal sleeve connects the feedwater sparger to thefeedwater nozzle. One end of the thermal sleeve is weldedto the sparger water inlet, and the other end is joined to thesafe end of the feedwater nozzle by a slip fit joint. A sim-plified sketch of the feedwater sparger to feedwater nozzle

CRDHOUSING

VESSELWALL

N/ J-WELD

STUB TUBE

11 -. .

Figure 5.2 Crack In CRD stub tube

31 31 ~~~~~~~NUREG/CR-5754

Survey

Table 5.1 BWR internals with reported SCC

Component Cases reported Failure location

Core spray sparger 6 (BWR-3, BWR-4) Circumferential cracks in sparger(distribution unknown) ring and at header arm to 'T' box

weld HAZsCRD housing 2 (1 BWR-2, 1 BWR-3) Stub lube at HAZ of CRD housing

to stub tube weldShroud head bolts 2 (2 BWR-4) Near root of collar to bolt weldAccess hole cover 1 (1 BWR-4) Access hole cover plate to shroud

support weldsCore spray line internal piping 2 (2 BWR-4) Weld HAZsSteam dryer support ring Several Cold-work locationsFeedwater sparger Several Weld HAZs in ejection nozzlesJet pump assembly holddown beam 6 (5 BWR-3, 1 BWR-4) Bearn ligament sections at beam bolt

diameterJet pump assembly riser pipe 2 (1 BWR-3, I BWR-4) Riser pipe to safe end weldsControl blades Several Sheath to tie rod and handle welds,

bail portion of handleIRM dry tube 4 (2 BWR-2, 1 BWR-3, I BWR-4) Top of thin tube segment surround-

ing compression springSRM dry tube 3 (2 BWR-2, I BWR-4) Top of thin tube segment surround-

ing compression springNeutron source holder Several Holder near top of beryllium

chamber

connection is shown in Fig. 5.3. In the original design, theslip fit joint is loose, and feedwater leaked into the narrowregion between the outside surface of the thermal sleeveand the inside surface of the feedwater nozzle. Instabilityof the leakage flow caused the feedwater sparger to un-dergo vibrations. In addition, the cooler leakage feedwateris mixed with the hot recirculating downcomer reactor wa-ter in the region surrounding the sparger water inlet. Flow-induced vibrations and rapid thermal cyclings led to thedevelopment of fatigue cracks in the feedwater sparger andaround the feedwater nozzle comer.

ORNL-DWG 92-2915 ETDREACTOR

The feedwater sparger fatigue cracking problem was re-solved by the implementation of a new sparger with a tightslip-fit joint, effectively reducing the leakage flow down toa very low and manageable level. A new feedwater nozzlewas also used to prevent crackings in the nozzle with thenew feedwater sparger design.

The BWR vendor has conducted extensive studies on thefeedwater sparger cracking problem, and the results aresummarized in an article by Kiss and Gerber.6

5.32.2 Jet Pump

Jet pumps in most early BWR-3 units were affected byflow-induced vibration problems, which led to the devel-opment of fatigue cracks in the pump support system. Thedesign deficiency was recognized early, and the fatiguecracking problem was resolved by using a new andstronger holddown beam. The improved jet pump supportwas implemented to many units before they become opera-tional.

Due to its operation characteristics, fatigue caused by flow-induced vibrations is always a concern with jet pumps inBWRs.

SLIP FITJOINT I

Figure 5.3 Feedwater sparger and feedwater nozzleconnection

NUREG/CR-5754 32

Survey

53.2.3 LPRM Dry Tube References

Indications of leakage from three Local Power RangeMonitor (LPRM) dry tubes were detected during a coldshutdown of an overseas BWR-6 unit. The reactor was op-erating in the low-pressure coolant injection (LPCI) mode.Damages to the LPRM dry tubes were confirmed in subse-quent inspections of reactor internals, and 46 pieces of abroken dry tube were retrieved. The damaged dry tube alsorepresented a degradation of a portion of the reactorcoolant pressure boundary. Failures were attributed to fa-tigue caused by flow-induced vibrations generated by theLPCI flow.

The problem was unique to BWR-6 units and was resolvedby the installation of a flow deflector at the LPCI dischargeopenings inside the shroud. The flow deflector shields theLPRM dry tubes from excitations generated by the LPCIflow.

The flow deflector was installed in domestic BWR-6 units.The reported failure data base showed no LPRM dry-tubefailure in domestic BWR-6 units.

Fatigue failures in other BWR internals were reported andare summarized in Table 5.2. Although some reactor inter-nals are susceptible to erosion, creep, and stress relaxationfailures, an extensive search of the LER data base did notlocate any reported failure in BWR internals that can beattributed to these particular aging-related degradationmechanisms.

1. American Society of Mechanical Engineers, ASMEBoiler and Pressure Vessel Code, Section XI, "Rulesfor Inservice Inspection of Nuclear Power PlantComponent."*

2. Science Application International Corp., "NuclearPlant In-Service Inspection Requirements andPractices in Different Countries: A ComparativeReview," Interim Report, EPRI NP-5919, July 1988.

3. "BWR Jet Pump Assembly Failure," IE Bulletin No.80-07, USNRC Office of Inspection and Enforcement,Washington, D.C., April 1980.t

4. "Cracking in Core Spray Sparger," IE BulletinNo. 80-13, NRC Office of Inspection andEnforcement, Washington, D.C., May 198 0.t

5. "Cracks in Shroud Support Access Hole CoverWelds," Information Notice No. 88-03, NRC Office ofNuclear Reactor Regulation, Washington, D.C.,February 1988.t

6. E. Kiss and T. L. Gerber, "Status of Boiling WaterReactor Structural Integrity Programs at the GeneralElectric Company," Nucl. Eng. and Des. 59, 27-45(1980).t

Avaulable from American National Standards Institute, 1430 Broadway,New York, NY 10018, Copyrighted.

tAvailabe in public technical libraries.

Table 5.2 BWR internals with reported fatigue failures

Component Cases reported Failure location

Steam dryer 8 (BWR-2, BWR4, BWR-5) Cracks in thin dryer hoods and lift-(distribution unknown) ing rod straps

Steam dryer support ring 1 (BWR4) Cracks in support bracketJet pump sensing line 5 (1 BWR-3,4 BWR-4) Attachment welds to diffusers and in

sensing linesJet pump restrainer gate 2 (2 BWR-3) Tack welds in outboard clamp bolt

keeperIn-core neutron flux monitor 1 (BWR-6) Cracks in dry tube at upper cooling

(LPRM) dry tube holesFeedwater sparger Several Circumferential cracks in sparger

near reactor pressure vessel feed-water nozzle

33 33 ~~~~~~~NUREG/CR-5754

6 ISI and Aging Management Program

Visual inspection is a relatively simple inspection methodused mainly to detect surface flaws and cracks. However,the method is limited to easily accessible regions of thecomponent being examined and is not capable of detectingsubsurface or partial through-the-wall cracks. Due to ac-cess limitations, interior regions of steam separators, OFS.pieces, the shroud head, and the core plate are not in-spected.

Alternate methods have been tried, on an experimental ba-sis, to detect partial through-the-wall cracks, as well asflaws in inaccessible regions in BWR internals. A methodthat has been used with some success is ultrasonic (IT) in-spection. A UT inspection has located partial through-the-wall cracks in the access hole cover that were not detectedby visual examinations; however, UT inspection has itsown access limitation problems, and the interpretations ofUT inspection results are more complicated and difficultthan are those for visual inspections. Many active researchworks are investigating alternate inspection methods; theobjective is to improve the overall effectiveness of theplant ISI programs. Establishment of a plant ISI program isan evolving process. When new information or new inspec-tion methods become available, the responsible organiza-tion will issue bulletins and letters to inform plant opera-tors of the latest developments and their implications in in-spection procedures and maintenance requirements.

One of the major concerns in reactor operations is the pres-ence of loose parts in the primary system. Because of thelong interval between inspections and the inability of vi-sual inspections to detect subsurface or partial through-the-wall cracks, the scenario that a reactor may operate withflawed or cracked internal components cannot be ruled out.The failure of reactor internals during operation may gen-erate loose parts that can cause damage to components inthe reactor primary system. Such failures have occurred inPWRs,1 but there is no reported incident of this nature inBWRs.

Loose parts can endanger the core support and core coolingfunctions of reactor internals. In addition to safety implica-tions, loose parts in the reactor primary system can lead toextensive outages for repair work. For safety reasons,'NRCis interested in the development of monitoring methodsthat can detect the presence of loose parts in the reactorprimary system. Research and development in detectingloose parts has led to the establishment of NRC RegulatoryGuide 1.133,2 which outlines the operating requirementsfor the establishment of Loose Part Monitoring Systems(LPMSs) in domestic commercial nuclear power plants;Reactors licensed since 1978 are equipped with LPMSs.

LPMSs have also been installed in many plants licensedbefore 1978.

6.1 LPMS

The LPMS is an acoustic-based monitoring system de-signed to detect the presence of loose metallic parts in thereactor primary system. It provides diagnostic informationto the plant operator so the operator can take appropriateactions to ensure a safe plant operation. These correctiveactions can also minimize the risks to other reactor compo-nents and systems.

Sound waves will be generated when a loose part, carriedalong by the reactor coolant flow, collides with a stationaryreactor component. The sound waves, basically bendingwaves, will propagate to other parts of the reactor, includ-ing the pressure vessel. The effectiveness of the LPMS willdepend on the system's capability to detect these impactwaves and the operator's ability to interpret these structure-borne signals.

The LPMS uses a series of sensors (piezoelectric ac-celerometers) mounted on the outside surface of the reactorpressure vessel to detect collision-generated, structure-borne sound waves. Four rings of three sensors each arerecommended for BWRs. One ring is located in the tophead region and another at the bottom head region of reac-tor vessels. The other two rings are mounted in the cylin-drical portion of the vessel, usually in the vicinity of outletnozzles. Inputs from sensors are filtered and then fed to amonitor that will record and analyze them. More detailedinformation on LPMSs can be found in the reports byKryter3 and Mayo.4 -

The impact sound waves contain information that can beused to estimate the mass and energy of the moving object,as well as the location of the point of impact. A dominantfrequency and the amplitude of the structure-borne soundwaves at a sensor location can be extracted from the inputsignals. Differences in arrival times at various sensor loca-tions are also measured. The mass and energy of the mov-ing part are estimated based on acomparision of the mea-sured dominant frequency and amplitude of the inputsound waves with a known calibration curve. The impactlocation is determined by using differences in arrival timesat known sensor locations and a triangulation process.Many uncertainties are associated with the signal process-ing procedures. As a general rule, the uncertainty level islow when the mass of the moving object is small and thepropagation path of the structure-borne sound waves is

35 35 ~~~~~~~NUREGICR-5754

Is'

simple. The uncertainty level increases with the mass ofthe loose part and the complexity of the sound wave propa-gation path.

The performance of LPMS is mixed. Because of the com-plexity and difficulty in processing and interpreting thestructure-borne sound waves generated in a collision pro-cess, mixed performance is not unexpected. EPRI has con-ducted a study with the goal of improving the performanceof LPMSs; recommendations from the study can be foundin the paper by Weiss and Mayo.5

While an LPMS can provide information that can be usedto detect the presence of loose parts in the reactor primarysystem, it lacks the capability of indicating the current sta-tus of system degradations that may exist in reactor com-ponents, structures, and systems. The safety of reactor op-erations and plant availability can be enriched by a moni-toring system that has the capability of providing informa-tion regarding the status of system degradations in a com-ponent. Based on the information, appropriate correctiveactions can be initiated before failures and malfunctionsoccur. Vibration monitoring and trending studies, whichare common practices in preventive maintenance programsfor rotating machineries, have potential applications in re-actor inspection and maintenance programs.

6.2 Vibration Monitoring andTrending Studies

The hostile environment inside a reactor pressure vesselwould rule out the use of accelerometers attached to inter-nals for long-term vibration measurements. Most reactorinternal vibrations are measured through indirect means. Acommon method is to infer internal component vibrationsfrom neutron noises measured by ex-core detectors.Neutron noises are fluctuations in the neutron flux arounda mean value. The neutron flux is moderated by the waterlayer between the core and the pressure vessel. Vibrations.of internal components located around the core can changethe thickness distribution of the water layer surroundingthe core. Changes in the water layer thickness can lead tovariations in the moderation effects of the water and can becorrelated to neutron noises as measured by ex-core detec-tors. Ex-core neutron noise measurement is considered aneffective method for detecting vibrations in internal com-ponents such as the core shroud, in-core neutron monitordry tubes, and fuel channel boxes.

The key to the neutron noise vibration measurementmethod is the establishment of a reference noise spectrum

at a given reactor power level and at specified sensor loca-tions. The noise spectrum is usually given in the form of aplot of the normalized power spectral density (NPSD)curve over a specified frequency range. The ability toidentify characteristic features or spikes in the noise spec-trum with structural natural frequencies of internals is es-sential to the success of the method. Neutron noise vibra-tion analysis is of limited value to internals whose naturalfrequencies cannot be clearly identified on the noise spec-trum. Fundamentals of the theory and practice of reactorneutron noise vibration measurement can be found in thetext by Thie.6

A reference noise spectrum can be obtained from results ofreactor preoperational testings. Temporary in-core sensorscan provide information that can be used to identify char-acteristic natural frequencies of internal components in thereference noise spectrum. The structural natural frequen-cies of reactor internals are also computed by analyticalmodels. Trending studies are performed by comparingnoise spectrums obtained during plant operations with thereference spectrum. Three to five neutron noise measure-ments are made during a fuel cycle. Deviations of an iden-tifiable point in an actual noise spectrum from that in thereference spectrum can be interpreted as indications ofdegradations in an internal component. The ability to corre-late deviations in the noise spectrums with the severity ofsystem degradation is essential to the success of trendingstudies.

It may be more difficult to establish an effective neutronnoise measurement system in BWRs than in PWRs. Bycomparision, the water layer thickness between the coreand the pressure vessel wall is larger in BWRs than inPWRs. The increase in water gap may reduce the sensitiv-ity of the water layer thickness vs changes in moderationeffects and make it more difficult to detect neutron noisesignals. Also, there are no existing ex-core neutron detec-tors in BWRs' As a result, neutron noise vibration monitor-ing and trending studies have not gained acceptance withBWR plant operators and reactor vendors.

Vibration monitoring and trending studies are used exten-sively in inspection and maintenance programs for Germanand French reactors,7 but their potential applications toplant ISI programs have not been fully explored in theUnited States.

The ISI program is an inspection and maintenance programthat is designed to detect failures before they can become athreat to the safety of plant operations. It offers little helpin controlling or reducing the effects of aging degradations.Aging-related degradation mechanisms can be managed by

NUREG/CR-5754 36

controlling or eliminating stressors associated with the ag-ing degradation mechanisms. Strategies used to manage thetwo major aging-related degradation mechanisms will beaddressed separately.

6.3 Strategy for Managing SCC

Reported aging-related failure information survey resultsshow that SCC is probably the most common aging degra-dation mechanism for BWR internals. Austenitic stainlesssteels, such as type 304, are used extensively in the con-struction of BWR internals. These steels provide goodgeneral corrosion resistance capability; however, the use ofwelding in the fabrication process and the reactor coolingwater may combine to create conditions that are conduciveto the development of SCC. As noted previously, threeconditions are needed for the development of SCC: a cor-rosive environment, a susceptible material, and the pres-ence of tensile stresses. All three conditions are present inBWR internals when weld HAZs come into contact withthe reactor cooling water.

The basic strategy for controlling SCC is the elimination ofany one of the three conditions. Material sensitization,which is responsible for the susceptibility of austeniticstainless steels, can be minimized by the use of stainlesssteels with a low carbon content and heat sink weldingmethods. However, for many existing operating reactorsthese are not viable options. The two remaining options forthese reactors are the reduction of the corrosive agent inthe coolant and the lowering of the tensile stress level in acomponent.

Tensile stresses in an internal component are caused by ex-ternally applied loads and by residual stresses in weldHAZs. Proper heat treatments can reduce residual stressesin weld HAZs, but they are not considered practical formost reactor internals because of the size of the compo-nents involved. When the dominant tensile stresses aregenerated by applied loads, the stress level in a componentcan be lowered by the use of larger structural components.Reducing the magnitude of the applied load is also effec-tive when it is possible to do so. These remedies have beenused to reduce the tensile stress level in jet pump holddown-beams. The reduction of the tensile stress level is an optionfor controlling SCC, but it has limited applications.

The remaining option for controlling SCC is to lower thedissolved oxygen content in the reactor cooling water. Thedissolved oxygen content can be reduced by the injectionof hydrogen gas into the coolant flow stream. The hydro-gen will react with oxygen to form water. The process be-

IsI

comes known as the Hydrogen Water Chemistry (HWC)8

program.

The dissolved oxygen content in the reactor cooling wateris a product of radiolytic reactions in the core and isgeneric to the reactor operation. A direct relationship existsbetween the dissolved oxygen content and the electrochem-ical potential (ECP) of austenitic stainless steels. When theECP is increased, the steel becomes more susceptible toSCC. The removal of the dissolved oxygen and loweringthe ECP to a value below -0.23 V (standard hydrogen elec-trode) will reduce the risk of the development of SCC ininternals made of austenitic stainless steels.

Hydrogen injection is an effective method to remove thedissolved oxygen from the coolant flow stream.Measurements in high-purity water showed a significantdrop in the ECP for austenitic stainless steels when the dis-solved oxygen content is reduced to below 40 ppb.9 Thesusceptibility to SCC for type 304 stainless steels is greatlyreduced when the dissolved oxygen content is kept below40 ppb. Laboratory testing resultsl 0 indicated that HWCcan inhibit the initiation of SCC, as well as retard thegrowth rate of existing flaws.

The implementation of a plant HWC program is an effec-tive method for controlling SCC in primary reactor coolantpiping systems. In principle, it should also be an effectivemethod for mitigating SCC in internal components. Theradiolytic reaction rate is higher around the core region,where most internals are located, and a higher hydrogengas injection rate may be required to reduce the local dis-solved oxygen content. However, field operating data withBWR internals in an HWC environment are very limited.As more plant HWC programs are implemented, a largeroperating data base will become available, and it can beused to assess the effectiveness of HWC for reactor inter-nals. The BWR vendor, GE, has issued a ServiceInformation Letter (SIL)1I to BWR plant operators rec-ommending the implementation of a HWC program in allBWR plants.'

Note that the implementation of a plant HWC program willincrease the radiation fields in the power-generating areassuch as the turbine building. The HWC program may comeinto conflict with other programs aiming at reducing radia-tion exposure levels to workers in nuclear power plants.

The presence of impurities such as chlorides and sulfatesmay also contribute to the development of SCC. Limits areimposed on the BWR water chemistry parameters to con-trol the impurity contents in the reactor cooling water. Theconductivity of water, which is a reliable indicator of theoverall level of impurities in the water, is monitored during

NUREG/CR-575437

ISIreactor operations. During normal BWR power operations,the operating specifications stipulate that the chloride andsulfate contents shall be maintained at 20 ppb or less, andthe water conductivity at <0.3 uS/cm.8 When the waterconductivity exceeds 0.3 gS/cm, the reactor water cleanupsystem is activated to lower the water conductivity to be-low its allowable limit. Prolonged operation in an off-stan-dard condition (in excess of 96 h) will require the imple-mentation of an approved program of corrective measuresthat will bring the level down below the allowable value.When the water conductivity exceeds 1.0 pS/cm for aperiod of 24 h, the reactor will be brought to a cold shut-down within 16 h. When the water conductivity exceeds5 gS/cm, the reactor will be shut down in an orderlymanner as rapidly as other plant constraints will allow. Thereactor water cleanup system and operating procedures willensure that the cooling water is maintained at an acceptablepurity level so that the degradating effects of impuritiescan be kept to a minimal level.

Other corrective actions can also be implemented to reducethe risk of SCC. The presence of crevices in ar, internalcomponent is a feature that could aid the development ofSCC, as well as reduce the effectiveness of HWC. Acrevice-free design and manufacturing process, togetherwith the use of stainless steels with a low carbon contentand heat sink welding method, could produce internalcomponents that are less susceptible to SCC. This com-bined approach should be used in making internals fornewer reactors and for replacement parts.

IASCC is a major aging-related degradation mechanismfor reactor internals located around the core. The effects ofa plant HWC program on the development of IASCC arenot well understood. A reduction of the corrosive agent inthe cooling water may reduce the risk of IASCC, but thereare not sufficient operating data to quantify the effects.

6.4 Strategy for Managing Fatigue

BWR internals are susceptible to high- and low-cycle fa-tigue failures. High-cycle fatigue is the more prevalentdegradation mechanism-..

The development of cracks in thin steam dry hoods is oneof the few cases of reported aging-related failures in whichlow-cycle fatigue was the suspected cause. The thin steamdry hoods were subjected to large structural deformationsunder cyclic pressure loads. As a result of these early prob-lems, thicker steam dryer hoods are used in the newer reac-tors and as replacement parts. Failures have not been de-tected in these new and thicker steam dryer hoods.

A second potential source for low-cycle fatigue failures inreactor internals is the development of flow-induced reso-nant vibrations. Reactor internal components are designedto prevent the development of large-amplitude flow-induced resonant vibrations; this is accomplished byseparating the component structural fundamental naturalfrequency from the dominant flow-induced excitation fre-quency. As an added measure, preoperation flow testingsare used to detect potential resonant and other types of vi-bration problems. The development of flow-induced reso-nant vibrations is a problem that has plagued the operationsof heat exchangers, but it is not a common problem for re-actor internals. In one reported case of fatigue failure, thedevelopment of flow-induced resonant vibrations may bethe suspected cause. This case involved a LPRM dry tubein a BWR-6 reactor. The reported failure was detected dur-ing a shutdown operation with the reactor in the low-pres-sure coolant injection (LPCI) operating mode. Damage wasattributed to fatigue failure at the dry tube upper coolingholes caused by FIV created by the LPCI flow. The prob-lem was resolved by the installation of a flow deflector toshield the LPRM dry tubes from flow-induced excitations.

No evidence suggests that the development of resonant vi-brations is a common problem for BWR internals.However, vessel internals are not immune to the effects offlow-induced small-amplitude vibrations, which could leadto high-cycle fatigue failures. The effects of small-ampli-tude vibrations are cumulative; depending on the stresslevel in the component, they could increase the probabilityof fatigue crack initiation in a component after a certainnumber of cycles of operation. For most stainless steelsand at specified stress levels, the allowable number of cy-cles of operation can be evaluated by using the fatiguecrack initiation curves specified in the ASME B&PVCode.12 The fatigue crack initiation curves, commonly re-ferred to as the S-N curves, tend to approach asymptotic -

stress values as the number of cycles increases. The S-Ncurves for some material level off beyond a certain numberof cycles of operation, and the corresponding asymptoticstress values are known as endurance limits. Theoretically,there is no limit to the number of cycles of operation whenthe maximum stress level in a component is kept below theendurance limit. Maintaining a low stress level in a reactorinternal component is the basic strategy for managinghigh-cycle fatigue problems. When the stress level cannotbe kept below the endurance limit or when the S-N curvesdo not level off, a counting procedure is used to keep trackof the number of cycles of operation; the information canbe used to estimate the time to fatigue crack initiation.Note that environmental factors have not been taken intoaccount in the development of the fatigue crack initiationcurves. Flaws in a component could also affect the time tofatigue crack initiation and alter the rate of crack propaga-tion.

NUREG/CR-5754 338

A fracture-mechanics analysis can be used to evaluate thetime interval between crack initiation and unstable crackpropagation. The rate of fatigue crack propagation is sensi-tive to environmental conditions, and attempts are beingmade to incorporate these factors into the analyses. At thepresent time, the confidence level in these predictive meth-ods is uncertain at best because environmental effects havenot been fully incorporated into the analysis.

IsM

6. J. A. Thie, "Power Reactor Noise:' American NuclearSociety, La Grange Park, III, 1981.

7. D. Wach, "Vibration, Neutron Noise and AcousticMonitoring in German LWRs, "Nucd. Eng. and Des.129 (1991).t

References8. "BWR Hydrogen Water Chemistry Guidelines: 1987

Revision," EPRI NP-4947-SR, Electric PowerResearch Institute, Palo Alto, Calif., December 1988.

1. "Characterization of the Performance of Major LWRComponents:' EPRI NP-5001, Electric PowerResearch Institute, January 1987.

2. U.S. Nuclear Regulatory Commission RegulatoryGuide 1.133, "Loose Parts Detection Program for thePrimary System of Light-Water Cooled Reactors."*

3. R. C. Kryter et al., "Loose Parts Monitoring: PresentStatus of the Technology, Its Implementation in U. S.Reactors and Some Recommendations for AchievingImproved Performance," Prog. Nucl. Energy 1, 2-4(1977).t

4. C. W. Mayo, "Loose Part Signal Theory," Prog. NucL.Energy 15 (19 8 5 ).t

5. J. M. Weiss and C. W. Mayo, "Recommendations forEffective Loose Parts Monitoring," Nucl. Eng. andDes. 129 (199 1).t

9. M. E. Indig and A. R. Mcllree, "High TemperatureElectrochemical Studies of the Stress Corrosion ofType 304 Stainless Steel," Corrosion 35,288 (1979).t

10. B. M. Gordan et al., "Hydrogen Water Chemistry,"Interim Report, General Electric Company, NEDE-30261, September 1983.

11. General Electric Company, Service Information LetterNo. 408, July 9, 1984.

12. American Society of Mechanical Engineers, "ASMEBoiler and Pressure Vessel Code, Section VIII,Pressure Vessels, Div. 1."*

Copies are available from U.S. Government Printing Office,Washington. D.C. 20402. A1rN: Regulatory Guide Account.

tAvailable in public technical libraries.*Available from American National Standards Institute, 1430 Broadway,

New York. NY 10018, Copyrighted.

39 NUREG/CR-5754

7 Discussions and Conclusions

BWR internals operate in an environment that is favorableto the development of aging-related degradation mecha-nisms. The primary stressors are a high-temperature, cor-rosive, and moving liquid coolant; manufacturing pro-cesses; fast neutron fluxes; and cyclic loadings. Reportedaging-related failures can be attributed to two major degra-dation mechanisms: SCC (including lASCC) and fatigue.The presence of crevices and cold-work locations in an in-ternal component may contribute to and accelerate the de-velopment of SCC problems.

7.1 SCC

The three conditions needed for the development of SCCare a corrosive environment, a susceptible material, andtensile stresses. Dissolved oxygen is the primary corrosiveagent in the reactor cooling water, and material sensitiza-tion can make austenitic stainless steels susceptible to cor-rosion attacks. Preloads in bolts and residual in weld HAZsare major sources of tensile stresses in an internal compo-nent.

The basis for formulating aging management strategies isthe control or elimination of any one of the conditions as-sociated with the development of aging-related degradationmechanisms. The use of stainless steels with a low carboncontent and heat sink welding methods can make internalcomponents less susceptible to the corrosive environment;however, these are not viable options for existing reactorsthat have not been built with these technologies. A popularapproach taken by the nuclear power industry to mitigateSCC in BWRs is to reduce the dissolved oxygen content inthe reactor cooling water. This is accomplished by the im-plementation of a plant HWC program in which hydrogengas is injected into the coolant flow stream to remove thedissolved oxygen. Testing results indicated that the plantHWC is effective in suppressing SCC in reactor primarycoolant piping systems. More operating data will beneeded to assess the effects of HWC on reactor internals.The use of crevice-free design and manufacturing methodsmay aid in the control of SCC in reactor internals in newreactors and in replacement parts.

The exposure to fast neutron fluxes is a major contributingfactor to the development of IASCC. IASCC has been de-tected in nonsensitized steel components with low stresslevels. By reducing the corrosiveness of the reactor coolingwater, the plant HWC program should be beneficial to thecontrol of IASCC. However, more operating data will beneeded to assess the effects of HWC in mitigating IASCCin reactor internals.

The implementation of a plant HWC program will increasethe radiation fields around the power-generating areas.This may limit the use of HWC because the increase in ra-diation level is in conflict with the goal of reducing radia-tion exposures to workers in nuclear power plants.

7.2 Fatigue

Reactor internals undergo vibrations as a response to dy-namic loads. Dynamic loads can be mechanical or thermalin nature. Vibrations will eventually lead to crack initiationand crack growth. The time to crack initiation is deter-mined by the levels of stress and structural deformations ina component.

Most fatigue failures in BWR internals are caused bysmall-amplitude FIVs. Unlike large-amplitude FIVs, it isnot possible to eliminate all small-amplitude vibrations,and high-cycle fatigue is an active aging degradationmechanism for reactor internals. Also uncertainties are as-sociated with the synergistic effects of fatigue and a corro-sive environment A vigorous inspection program isneeded to control fatigue cracking problems in reactor in-ternals.

7.3 Inspection Programs

When an aging-related degradation mechanism is not con-trolled or mitigated, it would eventually lead to failures inthe affected components. Effective inspection and agingmanagement programs are essential to the prevention ofaging-related failures.

BWR internals are visually inspected in accordance withthe rules and regulations of Sect. XI of the ASME B&PVCode. Visual inspections can detect surface flaws, but themethod has shortcomings. The major one is access limita-tion; namely, the inspection is only performed on parts thatcan be visually observed. As a result, internal componentswith a complex geometry or located in inaccessible regionsare not inspected on a regular basis. Another shortcomingof visual inspections is that they cannot detect subsurfaceor partial through-the wall cracks. Other inspection meth-ods, such as UT and eddy-current inspections, have beenused to examine internals. Both the UT and eddy-currentinspection methods have their own limitations. For thesereasons, the scenario in which a reactor may operate withflawed or cracked internal components cannot be ruled out.Reactors licensed since 1978 are equipped with LPMS.LPMS is an acoustic-based monitoring system and can de-tect the presence of loose parts in the reactor primary

NUREGICR-575441

Discussions

system. When a loose part is detected, corrective actionscan be taken to limit potential damages to other reactorcomponents and systems. LPMS does not possess thecapability of detecting incipient failures. Vibrationmonitoring and trending studies can provide suchcapability. Neutron noise measurement is a commonmethod for detecting vibrations in selected reactorinternals. Trending study is an assessment of the relativevibration amplitude of an internal component as a functionof time. Increases over a predetermined limit areinterpreted as signs for the incipient failure of acomponent. The large water gap between the core and thereactor vessel wall may reduce the effectiveness of neutronnoise vibration measurements for BWR internals. The lackof ex-core neutron flux monitors may also handicap the useof neutron noise vibration measurement in BWRs.

An effective vibration monitoring and trending studiesprogram can enhance reactor safety, as well as plant

efficiency. These preventive maintenance practices havebeen used sparingly but should warrant more considerationand exploitation by the U. S. utility industry.

Reactor internals, while they perform safety-related func-tions such as core cooling and core support, are not compo-nents of the reactor primary containment system. They canbe replaced if necessary. As a result, reactor internals arenot considered to be high-ranking safety-class components.Failures of internals could create conditions that maychallenge the integrity of the reactor primary containmentsystem, but they do not affect the effectiveness of theprimary containment systems. However, aging-relatedfailures may require extensive shutdown time for repairworks. The implementation of effective aging degradationmanagement and improved monitoring programs can im-prove the safety, as well as the efficiency, of long-term re-actor operations.

NUREG/CR-5754 442

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Internal Distribution

1. D. A. Casada 41. W. P. Poore III2. D. D. Cannon 42. C. E. Pugh3. R. D. Cheverton 43. J. S. Rayside4. D. F. Cox 44. T. L. Ryan5. C. P. Frew 45. C. C. Southmayd6. R. H. Greene 46. J. C. Walls7. H. D. Haynes 47. ORNL Patent Office8. J. E. Jones Jr. 48. Central Research Library9. R. C. Kryter 49. Document Reference Section

10. J. D. Kueck 50-51. Laboratory Records Department11-40. K. H. Luk 52. Laboratory Records (RC)

External Distribution

53. G. Sliter, Electric Power Research Institute, P.O. Box 10412, Palo Alto, CA 9430354. J. W. Tills, Institute for Nuclear Power Operations, 1100 Circle 75 Parkway, Atlanta, GA 30339-306455. R. J. Lofano, Brookhaven National Laboratory, Bldg. 130, Upton, NY 1197356. R. P. Allen, Battelle-PNL, MS P8-10, P.O. Box 999, Richland, WA 9953257. J. P. Vora, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Electrical and

Mechanical Engineering Branch, 5650 Nicholson Lane, Rockville, MD 2085258. G. H. Weidenhamer, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Electrical and

Mechanical Engineering Branch, 5650 Nicholson Lane, Rockville, MD 2085259. E. J. Brown, U.S. Nuclear Regulatory Commission, Office for Analysis and Evaluation of Operational Data,

Reactor Operation Analysis Branch, Maryland National Bank Building, 7735 Old Georgetown Road, Bethesda,MD 20814 C

60. C. Michelson, Advisory Committee on Reactor Safeguards, 20 Argonne Plasma Suite 365, Oak Ridge, TN 3783061. M. Vagins, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Electrical and

Mechanical Engineering Branch, Division of Engineering, 5650 Nicholson Lane, Rockville, MD 2085262. J. J. Bums, U.S. Nuclear Regulatory Commission, Division of Engineering Safety, Office of Nuclear Regulatory

Research, 5650 Nicholson Lane, Rockville, MD 2085263. M. J. Jacobus, Sandia National Laboratories, P.O. Box 5800, Division 6447, Albuquerque, NM 8718564. H. L. Magleby, Idaho National Engineering Laboratory, MS 2406, P.O. Box 1625, Idaho Falls, ID 8341565. V. N. Shah, Idaho National Engineering Laboratory, P.O. Box 1625, Idaho Falls, ID 8341566. Office of Assistant Manager for Energy Research and Development, Department of Energy, ORO, Oak Ridge, TN

3783167-68. Office of Scientific and Technical Information, P. O. Box 62, Oak Ridge, TN 37831

NUREG/CR-575443

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1 REPORT NUMBER12 891 SA.signed by NRC. Add Vol.. Suva.. REV..3201,202, BIBLIOGRAPHIC DATA SHEET NUREG/CR-5 7tL'

(See instructions on the reverse)UREG/CR-5754 lORNL/TM-1 18762. TITLE AND SUBTITLE

Boiling-Water Reactor Internals Aging Degradation Study 3. DATE REPORTPUBLISHEDMONTH | EAR

Phase 1 September 19934. FIN OR GRANT NUMBER

B08285. AUTHORIS) 6. TYPE OF REPORT

K. H. Luk7. PERIOD COVERED finclusiM Dates,

B. PERFORMING ORGANIZATION - NAME AND ADDRESS 1l NRC. providebjvision. Office or Region. U.S. Nuclear Regulatory Commission. and maihng address: ii contractor, providenats. nd mailing ag/ddrusr

Oak Ridge National LaboratoryOak Ridge, TN 37831-6285

9. SPONSORING ORGANIZATION -NAME AND ADDRESS 1li NRC, type "Same as aboee-: if contractor. peovide NRC ivision. Office or Region. U5. Nuclar aRegulatory Commission,and mailing address.)

Division of EngineeringOffice of Nuclear Regulatory ResearchUS Nuclear Regulatory CommissionWashington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words orilessThis report documents the results of an aging assessment study for boiling water reactor (BWR) internals.Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primarycoolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant,applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of acomponent failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the twomajor aging-related degradation mechanisms for BWR intemals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures weredetected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program isconsidered as a promising method for controlling SCC problems in BWR. More operating data are needed toevaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cyclefatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals areexamined by visual inspections and the method is access limited. The presence of a large water gap and anabsence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutronnoise vibration measurements, for BWR.

12. KEY WORDSIDESCRIPTORS (List words or pA roses that wvil assist rsearchrn in locating the teport.) 13. AVAI LA81 LIT Y STATEMENT

unlimitedNPAR, reactor internals, stressors, aging-related degradation 14.SECURITYCLASSIFICATION

mechanisms, component failure information, inservice inspections (ThiisParl

boiling-water reactor (BWR), stress corrosion cracking (SCC), fatigue unclassified(rhis Report)

unclassified15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (24891

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- -NUREG/CR-57,54' : ; -BOILING-WATER REACTOR INTERNALS AGING DEGRADATION STUDY, '- - -- :: : SPEBR19 - ,

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