NRC Confirmatory Testing and Analysis.

NRC ConfirmatoryTesting and Analysis

Stephen M. Bajorek, Ph. D.

Office of Nuclear Regulatory Research

United States Nuclear Regulatory Commission

Ph.: (301) 415-7574 / [email protected]

AP1000 Design Workshop

August, 2007

Slide 2

NRC Confirmatory Investigations

NRC performs independent analysis to confirm an

applicant’s conclusions.

NRC develops own safety codes to assist in review

and in identifying safety issues.

RELAP5

TRACE (RELAP5, TRAC-P, TRAC-B)

Experimental tests performed to develop data for NRC

code assessment & to support / confirm regulatory

decisions.

Slide 3

AP1000 Issues for NRC Codes

NRC performed extensive validation of RELAP5

for AP600 review.

Liquid entrainment issues also apply to NRC

developed codes:

Hot leg entrainment

Upper plenum entrainment

Calculations of AP1000 DEDVI break using

RELAP5 to confirm W NOTRUMP calculations.

Slide 4

NRC AP1000 Testing Progams

NRC sponsors work at several experimental

facilites to provide data for code development

and assessment. Applicable to AP1000 are:

APEX (Integral facility at Oregon State Univ.)

ATLATS (Separate effects test at Oregon State Univ.)

RBHT (Reflood heat transfer facility at Penn State Univ.)

Slide 5

Entrainment from

Upper Plenum Entrainment from Hot Leg

Jg3

ADS-4

Pool

Jg,H L

Small Cold Leg Break

Jg,UP

(minor)

(dominant)

Hot Leg Flow Pattern & Offtake to ADS-4

ATLATS simulates phase separation

at AP600/AP1000 branch line.

Slide 6

Conventional View ATLATS Test

ATLATS Experimental Investigation

ATLATS: Used to develop a database for hot leg entrainment and offtake at an upward facing branch line. Correlations for onset of entrainment and entrainment rate developed.

Slide 7

ATLATS Experimental Investigation

Slide 8

Entrainment from

Upper Plenum

Entrainment from

Hot Leg Stratified Layer

Jg3

ADS-4

Pool

Jg,H L

Jg,UP

DEG DVI Break

(dominant)

(minor)

Upper Plenum “Pool” Entrainment

Data from APEX-AP600 experiments

not well scaled for AP1000.

DEDVI calculations showed low water

levels in upper plenum.

Slide 9

APEX-AP1000 Facility

1. CMT - Larger Volume 6. PRHR Lines - Larger Diameter

2. ADS 4 - Larger Flow Nozzle Area 7. IRWST Injection Lines - 1 Line Larger Diameter

3. Pressurizer - Larger Volume 8. IRWST Liquid Level - Increased Height

4. PZR Surge Line - Reduced Diameter 9. Sump Recirculation Lines - 1 Line Large Diameter

5. Reactor - 1 MW Power and Larger Core Flow Area 10. Data Acquisition System Replacement

APEX facility

modified to more

closely represent

AP1000

Slide 10

APEX-AP1000 Facility Modifications

New PressurizerNew Core Make-Up Tanks (1 of 2)

Slide 11

APEX-AP1000 Facility Modifications

APEX Upper Internals

Upper Core

Plate

Upper

Internal

Assembly

Guide Tube Details

Slide 12

APEX-AP1000 Facility Modifications

New Data Acquisition System (DAS)

Slide 13

APEX-AP1000 Testing

Westinghouse needed to obtain new data to resolve liquid entrainment issue.

Tests began in the newly modified “APEX-AP1000” facility in March 2003. Modifications were made through DOE-NERI grant funding.

A total of 11 tests completed in 2003-2004, jointly funded by U.S.DOE and NRC. (Data was shared, but evaluated independently.) RES has evaluated 7 tests considered most important to help resolve questions on entrainment and facility scaling.

Slide 14

APEX-AP1000 Testing

Testing in the APEX-AP1000 facility

concentrated on the Double-Ended Direct

Vessel Injection (DEDVI) line break. This was

found to be most limiting break in AP600.

Two type of tests conducted:

Design basis: Single failure ( DOE / W )

Beyond design basis: Multiple failures ( NRC )

Slide 15

AP1000 Licensing Significant Tests

DEDVI with ¾ ADS4 valves available. One ADS4 valve on

pressurizer side failed closed.

DBA-03

DEDVI with ¾ ADS4 valves available. One ADS4 valve on

non-pressurizer side failed closed.

DBA-02

2-inch CL break with 2/4 ADS4 valves available. Both ADS4

valves on non-pressurizer side failed closed.

NRC-AP1000-06

DEDVI with 2/4 ADS4 valves available. Both ADS4 valves on

non-pressurizer side failed closed.

NRC-AP1000-05

1-inch CL break with degraded sump.NRC-AP1000-04

DEDVI with 2/4 ADS4 valves available. Both ADS4 valves on

pressurizer side failed closed.

NRC-AP1000-03

DEDVI with complete failure of ADS-1/2/3.NRC-AP1000-01

Test Number Test Description

Slide 16

Design Basis Test Results(Test DBA-03)

0 200 400 600 800 1000 1200 1400 1600 1800 2000Time (s)

40

50

60

70

80

90

100

Tw

o-P

has

e L

evel

(in

)

LT-120

DBA-03 (OSU-AP1000-04)DEDVI line break with pressurizer side ADS-4 valve failure

Bottom of Hot Leg

Top of Core

Westinghouse Proprietary Class II

Slide 17

Design Basis Test Results(Test DBA-03)

0 200 400 600 800 1000 1200 1400 1600 1800 2000Time (s)

0

5

10

15

20

Volu

metr

ic F

low

Rate

(gpm

)

FMM-504FMM-702FMM-402

DBA-03 (OSU-AP1000-04)DEDVI line break with pressurizer side ADS-4 valve failure

ACCUM

CMT IRWST

Westinghouse Proprietary Class II

Slide 18

Design Basis & Beyond Design Basis

Design Basis Event Testing

Examines plant response for a hypothetical accident with single

most limiting failure.

Consistent with “Chapter 15” analyses, and helps confirm margin

to regulatory limits.

Beyond Design Basis Event Testing

Examines plant response for a hypothetical accident with multiple

failures.

Help confirm PRA assumptions.

Can identify how much additional margin is available.

Provides data for code assessment / development for conditions

that probably do not occur in most DBA scenarios.

Slide 19

Beyond Design Basis Test Results(Test NRC-AP1000-05)

0 2 00 4 00 6 00 8 00 10 00 12 00 14 00 16 00 18 00Time (s )

40

50

60

70

80

90

10 0

Tw

o-p

hase

level, in.

L T-120

NRC-AP1000-05DE DVI with b oth non-p zr s ide A DS4 valv es failed closed

HL

UCP

ADS4-2 opens

CMT-2 empty

ACC-2 empty Heatup

Westinghouse Proprietary Class II

Slide 20

Beyond Design Basis Test Results(Test NRC-AP1000-05)

0 200 400 600 800 10 00 12 00 14 00 16 00 18 00Time (s )

0

0. 2

0. 4

0. 6

0.8

1A

DS4-2

Flo

w Q

uality

NRC-AP1000-05DEDV I with bo th n on-pressurizer side ADS4 v alves failed

CMT-2 empty

Z2 at UCP level

Heatup begins

Westinghouse Proprietary Class II

Slide 21

APEX-AP1000 Experimental Observations

Design-basis tests showed:

No core uncovery or cladding heatup observed; two-phase levels near or above bottom of hot leg.

Higher entrainment than AP600 tests. (DBA-02 vs NRC-20)

Less margin to core uncovery than in AP600. (DBA-02 vs NRC-20)

Sensitivity to ADS4 valve location failure. (DBA-02 vs DBA-03)

Beyond design basis tests:

Showed failure of 2/4 ADS4 valves causes core uncovery. (NRC-05 for DEDVI and NRC-06 for 2-in CL break)

Sensitivity to ADS4 valve location failure. (NRC-03 vs NRC-05)

Entrainment to ADS4 continues even when UP two-phase level drops to UCP. (NRC-05 and NRC-06)

Confirmed “robustness” in AP1000 design. Multiple ADS failures in specific locations needed to produce core uncovery.

Slide 22

Use of Experimental Findings in Review

Data from APEX-AP1000 were used to identify code

deficiencies and necessary corrections to Evaluation

Model.

APEX-AP1000 data

confirmed large

safety margin in

AP1000 design.

No uncovery /

cladding heatup for

design basis events.Westinghouse Proprietary Class II

Slide 23

LOCA Long Term Cooling

SG

IRWST

RECIRC

CONTAINMENT

PZR

ADS 4

PXS VALVEROOM

ADS 1/2/3

SPARGER

RCP

GUTTER

WASTESUMP

VESSEL

RECIRCSCREEN

CORE

STEEL

Slide 24

Core Collapsed Level (CLL) During

Long Term Cooling (LTC)

WCOBRA/TRAC long-term cooling calculation

predicted core collapsed level (CLL) at approx.

50% of core height with no cladding heatup.

Can heatup occur at this CLL ?

NRC performed independent calculations of

mixture level swell and also showed no heatup

with similar CLL.

Conclusion supported by NRC sponsored test program.

Slide 25

Mixture Level Swell

Rod bundle data for mixture level swell at low pressure, P < 3.44

bar (P < 50 psia), are relatively scarce.

Data most commonly used for code assessment does not have

correct range of inlet subcooling for AP1000. Power shape of

interest is top-skewed, rather than the chopped-cosine shape

used in most tests.

Rod Bundle Heat Transfer (RBHT) “interfacial drag” tests

recently completed provide additional information. RBHT

power shape is peaked near top of bundle.

Slide 26

Rod Bundle Heat Transfer (RBHT) Facility

The RBHT 7x7 rod bundle is

well-instrumented, with 45 full-

height (12-ft.) rods. Testing in

2003 included matrix of 44

“interfacial drag” tests to provide

detailed measurements of axial

void fraction.

Instrumentation includes steam

probe rakes and a laser

illuminated digital imaging

system to record droplet size.

Slide 27

Rod Bundle Heat Transfer (RBHT) Facility

0 0.2 0.4 0.6 0 .8 1

Elevation

0

0. 5

1

1. 5

2

Rel

ativ

e P

ow

er

A P1 000RBHT

RBHT power shape linearly increases to peak near 9-ft.

RBHT tests conducted at:

20, 30, 40 psia (0.138, 0.207, 0.276 MPa)

0.15 < VIN < 1.0 in/s

(0.381 < VIN < 2.54 cm/s)

20 oF < DTsub < 100 oF

(11 < DTsub < 56 K)

0.1 < ALHR < 0.26 kW/ft

Slide 28

RBHT Uncovery Results

0 10 20 30 40 50 60

Core Collapsed Level (%)

0

10

20

30

40

50

60

70

Core

Unco

very

(%

)

RBHT 1648RBHT 1637

THETIS Boil DownTHETIS Boil Up (30%)

THETIS Boil Up (40%)

THETIS Boil Up (50%)

RBHT 1648

P = 30 psia

ALHR=0.1 kW/ft

VIN=0.2 in/sec

DTsub= 40 oF

RBHT 1637

P = 30 psia

ALHR=0.1 kW/ft

VIN=0.2 in/sec

DTsub= 20 oF

Slide 29

Summary

NRC performed independent assessments to validate

Westinghouse conclusions on AP1000 performance.

NRC calculations support the conclusion that AP1000

will not have core uncovery or cladding heatup for

design basis small break LOCAs.

Test data from NRC sponsored facilities were used for

code assessment and resolution of review issues.