ET-NRC-93-4029, Non-Proprietary Presentation Matl from 931210 Meeting Between Westinghouse & NRC on AP600 Test Program

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Non-Proprietary Presentation Matl from 931210 Meeting Between Westinghouse & NRC on AP600 Test Program
ML20058Q400
Person / Time
Site: 05200003
Issue date: 12/10/1993
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19311B279 List:
References
ET-NRC-93-4029, NUDOCS 9312290075
Download: ML20058Q400 (129)
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Text

1__. .za Attachment 3 to Westinghouse Letter ET-NRC-93-4029 Presentation Material from December 10,1993 Meeting between Westinghouse and NRC on AP600 Test Program l

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9312290075 931216 PDR ADOCK 05200003 >

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' Yoa 0sv Westinghouse /NRC Test Issues Meeting

- l December 10,1993 OWFN 1-F-19 (8:30 - 4:00)

1. Introduction B. A. McIntyre ,
2. Core Makeup Tank Tests L. E. Hochreiter Westinghouse
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung ATWS A. C. Cheung
4. OSU Tests Higher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica
5. Noncondensable gas issues L. E. Hochreiter
6. PR.HR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closeout Summary of Remaining Issues & Actions All L ___ . _ _ - - . . . -.

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AP600 TESIPROGRAM ISSUES j I

CORE MAKEUP TANK (CMT) TESTS j TEST MATRIX: l "THE CMT MUST PERFORM ITS FUNCTION OF DELIVERING i WATER TO THE DOWNCOMER A) WHILE REACTOR COOLANT'  !

CIRCULATES FROM THE COLD LEG TO THE UPPER LEVEL OF- l THE CMT, B) WHILE A MIXTURE OF HYDROGEN AND STEAM  !

FLOWS FROM THE PRESSURIZER TO THE UPPER LEVEL OF THE CMT, AND C) WHILE PRESSURIZED COLD WATER AND THEN. i NITROGEN FLOW FROM AN ACCUMULATOR INTO THE'SAME j DELIVERY LINE THAT CONNECTS THE CMT TO THE i

- DOWNCOMER. THE CURRENT TEST MATRIX DOES NOT- l ADDRESS THE COMPLEX BEHAVIORS CONCOMITANT WITH THESE INTERACTIONS. THE STAFF HAS SUGGESTED CHANGES ' ,

IN THE TEST MATRIX THAT HAVE NOT BEEN MADE."

l SCALING ANALYSIS:

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'NONE HAS BEEN PROVIDED IN RESPONSE TO THE STAFF'S REQUEST".  ;

i 0672U4120093  !

. ... j CMT TEST UPDATE

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a ADDITIONAL INSTRUMENTATION HAS BEEN ADDED TO  !

ADDRESS NRC/INEL QUESTIONS l i

REVISED CMT TEST MATRIX ADDRESSES NRC CONCERNS IN  !

NOVEMBER 4TH LETTER  !

i CMT SCAUNG LOGIC REPORT -t i

UPCOMING W/NRC MEETING ON THE CMT TESTS

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REVISED CMT INSTRUMENTATION 8 NEW FLUID THERMOCOUPLES HAVE BEEN ADDED TO BETTER DETERMINE THE MDONG EFFECTS AT THE TOP OF THE CMT, TOTAL OF 10 THERMOCOUPLES HAVE BEEN ADDED

+

4 PRESSURE DROP CELLS HAVE BEEN ADDED OR THEIR ATTACHMENTS HAVE BEEN REVISED TO BETTER CHARACTERIZE THE CMT STEAM FLOWS AND LINE PRESSURE DROPS A LARGER DEPRESSION VALVE HAS BEEN ADDED TWO STEAMFLOW METERS HAVE BEEN ORDERED TO HELP MEASURE BALANCE LINE STEAM FLOWS 0 6 120093

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l CORE MAKEUP TANK TEST j l

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l PRE-MATRIX TESTING l l

5 COLD PRE-OPERATIONAL TESTS; OVER 100 EXPERIMENTS l

1 5 HOT PRE-OPERATIONAL TESTS; NEARLY 300 EXPERIMENTS  !

i EVALUATED STEAM DISTRIBUTOR DESIGN l i

REMAINING PRE-MATRIX TESTING l v

CHECKOUT OF LEVEL INSTRUMENT AND DAS ONGOING  :

i DEPRESSURIZATION SYSTEM CHECKOUT AND VALVE  !

CHARACTERIZATION f

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. e CMT TEST MATRIX SPECIFIC QUESTIONS FROM NRC IN NOVEMBER 4TH LETTER ,

ARE ADDRESSED

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FLOW RANGES AND PRESSURES ARE SIMILAR TO, BUT NOT EXACTLY THOSE RECOMMENDED BY INEL

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ADDITIONAL EMPHASE ON CMT RECIRCULATION BEHAVIOR 1

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l 62 MATRIX TESTS WITH STEAM DISTRIBUTOR INSTALLED ARE PLANNED TO ADDRESS:

WALL CONDENSATION WITH AND WITHOUT NON-CONDENSIBLES SURFACE CONDENSATION CONSTANT DRAINDOWN i

DRAINDOWN DURING DEPRESSURIZATION NATURAL CIRCULATION FOLLOWED BY DRAINDOWN AND DEPRESSURIZATION CMT ACTUATION WITH BOTH STEAM LINES, WITH AND WITHOUT NON-CONDENSIBLES 0572LH-120893

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1 CMT SCALING LOGIC REPORT l i

i AGREED ON KEY THERMAL-HYDRAULIC PHENOMENA FOR CMT BEHAVIOR WITH NRC l

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i DEVELOPED A PIRT TABLE FOR CMT PHENOMENA l

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DEVELOPED THE SCALING RATIONAL FOR:

CMT RECIRCULATION CMT DRAINDOWN 0672LH 120093 3

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KEY RESULTS FROM RECIRCULATION ANALYSIS l 4

RECIRCULATING MASS FLUXES ARE SIMILAR BETWEEN CMT TEST AND PLANT DEVELOPMENT OF HOT LIQUID LAYER IS SIMILAR BETWEEN CMT TEST AND PLANT A SMALL DISTORTION DOES E)0ST DUE TO STEAM / WATER RESERVOIR COLD LEVEL BUILDUP OC72LH 120093

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Figure 2-6 Recirculation Ratio of the CMT Test to the AP600 CMT at 1100 psia WPF1877D:Id/120893 38 l

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1 Figure 2 7 Comparison of the Hot Layer Thickness of the CMT Test and the Plant CMT at 1100 psia t

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CMT DRAIN DOWN ANALYSIS PERFORMED CALCULATIONS AT DIFFERENT CMT LEVELS AND  !

DIFFERENT ASSUMED MIMNG DEPTHS i

CALCULATED n VALUES FOR WALL AND SURFACE j CONSERVATISM .-  ;

CALCULATED n RATIOS TO EXAMINE SCALABILIT( i RESULTS INDICATE THAT WALL CONDENSATION IS SCALED AT BEGINNING OF TRANSIENT (500 SEC.) THEN IS LESS DUE TO ,

WALL THICKNESS i

SURFACE CONDENSATION TC RATIO'S VARIES FROM 0.25 TO 1.4 i i

THEREFORE, CMT TEST WILL ADDRESS KEY PHENOMENA i

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Figure 3-43 Calculated Wall Condensation Il Group Ratio at a CMT Level of 95% for Different Mixing Depths WPFI877D-3:1D/I20793 120

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ll Figure 3-46 Calculated Surface Condensation fl Ratio for CMT Level ef 95% and Different i Mixing Depths WPF1877D-3:lD/120793 123

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._WjNRC MEETING ON CMT TESTS AT W i

  • PROPOSE A MEETING THE WEEK OF MARCH 14TH,1994 t

ACRS HAS REQUESTED SIMILAR MEETING 1

PROPOSE TO HAVE BACK-TO-BACK MEETING SIMILAR TO OSU 1

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1 0672LH-120893

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CONCLUSIONS:  ;

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l TEST MATRIX HAS BEEN REVISED TO INCORPORATE NRC  !

CONCERNS FROM NOVEMBER 14TH LETTER i INSTRUMENTATION ADDITIONS HAVE BEEN ADDED AS PER NRC/lNEL RECOMMENDATIONS AND W NEEDS l l

PRELIMINARY SCALING LOGIC REPORT GIVEN TO NRC ON l 12/10/93 MEETING TO BE SCHEDULED IN EARLY MARCH FOR FURTHER DISCUSSION OF CMT TEST / ANALYSIS L

ACTIONS COMPLETE ITEMS IDENTIFIED IN NOVEMBER 14 LETTER 1

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NRC Request Wastinghouse Response Scaling Rationale  !

e Provide a scaling rationale for the A preliminary ecpy of the CMT [

CMT ter. facility. Scaling Report has been completed  !

and will be transmitted to the staff l for review.

I Consider explicit specific thermal Specific thermal hydraulic behavier '

hydraulic behavior in the scaling is delineated in the PIRT clart in <

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raticnale that may occur as subsets the scaling rationale for the CMT ,

! of the general " key phenomena". test.  ;

Explicitly show how the CMT tests The " key phenomena" is directly  :

relate to " key phenomena", related to the CMT test matrix and' delineated in the Scaling Report.  !

i Computer Code Models  !

)~ Inform the staff when any changes Any model changes made to the SSAR are made to any and all computer codes will be identified in the test code raodels as a result of the CMT analysis report on the CMT tests.

testing program.  !

Assess the ability of the SSAR codes The Test Analysis Report will  !

to properly represent the thermal contain a code-to-data comparison of '

hydraulic behavior which occurs in the phenomena exhibited in the the CMT. performance of the CMT test. ,

Examine the existing models in the The Test Analysis Report will SSAR codes to determine if they are contain a code-to-data comparison of ap.dicable over the ranges of the phenomena exhibited in the p inent parameters- that will performance of the CMT test.

e: Jt in the AP600 with respect to '

thermal hydraulic behavior which c ars in the CMT.

Facility Description Report Provide a facility cescription A Facility Description Report is in ,

report including as-built dimensions preparation for the CMT test which ,

and loss coefficient s. will include the design basis for- l the facility, as-built' drawings,.  !

piping and instrumentation drawings, i loss coefficients and other relevant )

information. A Quick-Look Report 'l has been prepared delineating the l results of the CMT cold pre-  ;

operational tests, which also ,

contains the piping loss j coe f ficie nt s.

Test Specification

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Provide an updated test An updated CMT Test Specification specification including an updated has been prepared and is under-test matrix. - review, and will be transmitted to the staff. _ An updated test matrix is included in the specification.

Include an additional objective in The objective to assess the the test specification to assess the performance of a prototypic type performance of the heated- level instrument is included in the thermocouple level sensors that will updated specification. This be used to trigger ADS in the AP600, instrument is installed in the CMT test facility.

Test Matrix Add tests with hot water volumes Tests 45 and 46 will be performed comprising between about 10% and 30% with hot water comprising 20% of the of the CMT prior to draindown and tank volume prior to draindown and.

depressurization. depressurization; tests 47 and 48 at 50% and tests 49 and'50 at 100%.

Simulate the presence of non- Tests 06 and 07 will be performed to condensible gases as closely as evaluate the effect of nitrogen on practical in the test facility. CMT wall condensation. Tests 08, 09, 10, 52, 53, 55, 56, 58, 59, 61 and 62 will examine the effect of helium (simulating hydrogen) on wall condensation and CMT actuation.

Depressurization rates need to be The CMT facility has been modified well specified. The rates should be to facilitate a wide range of chosen carefully, depressurization rates. -These rates will be specified based on the results of the hot pre-operational tests to be performed to characterize the facility depressurization system and the SSAR analysis.

Tests should be conducted at 20, Tests will be conducted at.10, 135, 500, 1100, and 2250 psia. 685, 1085 and 2235 psig.- These pressures were selected to provide a bread range of pressures, distributed saturation temperatures, and represent the pressures at which CMT draining will occur over the range of anticipated transients.

[Use) CMT drain rates of CMT drain rates of 6, 11, 16 and approximately 5.7, 11.1, 16.2 and 25 maximum flow will be used for gpm to bracket those expected in the draindown tests conducted in the CMT plant, test facility. The maximum flow is about 28 gpm.

[ Surface condensation) tests should Tests 11 through 18 will be be performed at steam volume heights performed at CMT steam heights of of 11.1 (10%) and 40.3 [35%) inches. 10%, .2 5 % , 50% and 75%.

(For draindown tests followed by Discharge line resistances will be depressurizaticn], reduce the set to achieve drain rates of 6 and maximum draindown rate to those with 16 gpm in these tests.

a consistent scaling approach.

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Conduct tests to cbserve the effect Tests 01 through 10 will be i of depressuri=ation on condensation conducted using Steam Line No. 1, ,

rates using Steam. Line No. 1 only. and will be depressurized to observe

, the effect on condensation rates.

Test Operating Procedures Provide the staff with a detailed Test Operating Procedures have been set of operating procedures written for the Pre-Operational Cold j and Het CMT tests. These procedures '

will be part of the Westinghouse i test records and are available for  ;

staff review. These procedures will be provided for staff review at the .

CMT test facility during the test j program. i l

Additional Instrumentation I Several thermocouples should be Additional thermocouples were ,

installed in the area of the CMT installed in the CMT dome between dome. elevations 110.5 and 114.5 at one- '

half inch to one inch intervals. '

(T/C No's 62, 78, 79, 63, 80 and 81) l Add instrumentation to measure steam Steam Line #1 has two pressure '

flow rates (e . g . , turbine ficw transducers installed (PDT 8 and meters) in Steam Lines #1 amd #2. 11), Steam Line #2 also has two pressure transducers (PDT 9 and 12) . 1 A fifth transducer is common to both lines (PDT 10) . These measurements are used with the measured piping loss coefficients to determine steam flow rates. Steam turbine meters are being obtained to provide a direct and redundant steam flow measurements in each line.

1 J two fluid thermocouples at the Thermocouples have been added at 3/4 full elevation to aid in elevations 104.1 (9/10 height) and determining the shape of the radial 56.6 (1/2 height) at radial fluid temperature profile. positions 6 and 9 inches from the-inside tank wall. (T/C No's 82, 83, 84 and 85)

Additional differential pressure Differential pressure measurements measurements are recommended: were considered sufficient for the test facility and were not added as 1-SWR outlet to SLf2 isolation valve suggested. However, during our inlet review of these measurements, three 2-Across SL42 isolation valve additional needs were identified:

3-SL#2 isolation valve to SL#1 tee 4-Across SL#1 isolation valve 1-SWR outlet to CMT diffuser inlet 5-Stil tee to top of CMT 2-CMT diffuser inlet to top of CMT 6-CMT outlet to SWR inlet 3-SWR outlet to SLil isolation valve inlet 4

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Westinghouse /NRC Test Issues Meeting December 10.1993 OWEN 1-F-19 (8:30 - 4:00)

1. Introduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Westinghoure Station Blackout A. C. Cheung ATWS A. C. Cheung  !
4. OSU Tests Higher Risk Shutdown event T. L. Schulz i

Capability to model SGTR and MSLB E. J. Piplica

5. Noncondensable gas issues L. E. Hochreiter 1
6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz j
8. Meeting Closeout Summary of Remaining Issues & Actions All l

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TEST ISSUE DISCUSSION Topic $

SPES-2 Multiple SGTR with ADS !

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i SPES-2 Multiple SGTR with ADS I

r Background - Westinghouse's understanding of the issue and the Staff's concern.

. In SRM for SECY 93-087, the commission approved: '

Multiple SGTR (2 to 5 tubes) be included in application for ,

design certification for passite PWRs, but '

Multiple SGTR is beyond design basis requirements for PWRs.

Realistic or best estimate analytical assumptions may be used to assess plant response.

. Staff has taken position that multiple "G FR be tested in SPES 2.

In response, E included a simulated 3 tube break (between 2 to 5 tubes) in SPES-2 matrix.

Originally, Staff recommended that rupture up to 5 tubes be considered.

. Since then, ACRS suggested a 7 tube rupture based on hexagonal configuration of AP600 Steam Generator

- Staff also recommends the multiple SGTR be tested with actuation of ,

ADS j

- Concern that this event will result in unexpected plant behavior.

l 4

. Since E analyses show that ADS will not be activated automatically j during a single or multiple SGTR (up to 10 tubes), it is assumed that l the staff is concerned with manual actuation.  ;

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W Evaluation ofIssue Automatic ADS Actuation t

AP600 SSAR Chapter 15.6 included the design basis SGTR (single tube) accident.

- Use a modified version of NRC approved LOFTTR2 Methodology and assumptions used consistent with those accepted for NTOL plants except no operator recoveryactions niodeled. -

i

= Results Demonstrated Acceptable radiological consequences per SRP No SG overfill predicted No ADS actuation

- SPES-II included two design basis single tube rupture tests Will provide adequate data to validate the computer code for design basis accident analyses f

For the beyond design basis multiple SGTR event analysis, W uses the MAAP4 code to assess the plant response. ,

Consistent with commission position of using more realistic /best estimate assumptions on SECY-93-087

- Cases of 2 to 10 tube ruptures analyzed i

- No ADS actuation predicted '

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1 SSAR RESULTS AP60] STEAM GENERATOR TUBE RL'.URE l RCS PRESSURE I

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SSAR RESULTS

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! AP600 STEAM GENERATOR TUBE RUPTURE i

i CORE MAKEUP TANK WATER VOLUME

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W Evaluation ofIssue Automatic ADS Actuation (Cont) i Since SMAR submittal, W has also performed realistic /best estimate analysis of a a iabe rupture with the small break evaluation tool NOTRUMP. '

Response to RAI 440 27

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LOFTTR2 is not designed or suitable for transients where ,

significant 2 phase flow and phase separation is expected.

- Provides an independent benchmark of MAAP4 calculations  !

- Best Estimate Approach Assumptions

Startup feedwater modeled (200 gpm/SG)

One PRHR actuates on high SG 1 secondary level (nominal value) '

No charging flow modeled t

ANS 1979 decay heat with no uncertainty Nominal reactor trip,"S" setpoints No additional failures assumed Break at cold side of tubesheet (DEG of each tube)

No operator actions assumed

. The NOTRUMP 5 tube rupture results show: ,

General plant response for a 5-tube rupture is similar to design basis single tube rupture i

AP600 plant response is similar to current plant with  :

respect to relationship between single and multiple SGTR ADS is not predicted to actuate  !

Only a slight loss of CMT level I

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PRELIMINARY NOTRUMP 5 TUBES RUPTURE Primary and Secondory System Pressure Primary Pressure c aBroken SG Pressure

. - Intact SG Pressure '

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PRELIMINARY NOTRUMP S TUBES RUPTURE Primary to Secondary Break Flow Liquid Greak Flow c o Vopor Breck Flow 300 e

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Core Uakeup Tank LeveI Groken Loop CMT I u a intoet Loop CMT  ;

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l W Evaluation ofIssue j Automatic ADS Actuation (Cont)

Based on the NOTRUMP and MAAP results, W concluded that multiple tube rupture (up to 10 tubes):

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1 Will not cause ADS to actuate Will not result in unexpected or totally different plant behavior from the design basis (single tube) accident  :

1 Is not needed for validation of the LOFTTR2 code, and l 1

Any phenomena occurring during the multiple SGTR should be captured by the 3 tube test proposed for SPES 2 l 1

The NOTRUMP 5-tube rupture calculation will be submitted to staff by end j of January 1994 l

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W Evaluation ofIssue Manual ADS Actuation

  • The expecte.d operator response to a SGTR (single or multiple) event will include the following actions:

- Identify faulted SG with rupture

    • Radiation alarm SG Ievel deviation

- Isolate faulted SG

    • Steam and feed flow

- Cool down RCS to desired subcooling using intact SG

" Steam dump to condenser if available

" Atmospheric dump Depressurize RCS to rupture SG pressure Pressurizer normal spray

    • Auxiliary spray Control depressurization with 1st stage valve Bring plant to and cold shutdown conditions Normal RHR

= Given the ability to clearly identify a single (or multiple) SGTR, the -  !

operators will not be directed to, and are not expected to, actuate the ADS l manually. i Has highly reliable equipment available for recovery under realistic assumptions.

Does not have to take any recovery actions to maintain plant in a safe condition under design basis assumptions.

I

- Operators will be directed to actuate ADS manually only if an automatic actuation signal has been generated

- Since manual actuation of ADS during single or multiple tube rupture is not likely or expected.

- Adding this test (SGTR with ADS due to manual actuation) is not warranted.

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SUMMARY

i W has shown that there is no automatic ADS actuation for single or multiple l (up to 10 tubes) SGTR based on realistic /best estimate calculations.

Phenomena occurring during multiple SGTR captured by the proposed 3-tube test at SPES 2 Data from SPES 2 SGTR tests is adequate to validate safety analysis codes l Operators are not directed to manually actuate ADS during recovery from single or multiple SGTR.

Changing the proposed matrix test to cover more than 3 tubes and adding test to look at SGTR with ADS will cause unnecessary delay in testing schedule and add cost without justifiable benefit.

i l

l RECOMMENDATION I i

W recommends that the proposed SPES test matrix for SGTR (2 design basis  ;

cases and one 3 tube rupture case) be accepted without change.

I l

Westinghouse /NRC Test Issues Meeting l

1 December 10,1993 OWFN 1-F-19 (8:30 - 4:00) l

1. Introduction B. A. McIntyre l
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung Westinghouse I

- ATWS A. C. Cheung

4. OSU Tests

)

l Higher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica

5. Noncondensable gas issues L. E. Hochreiter
6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closeout Summary of Remaining Issues & Actions All l 1

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TEST ISSUE DISCUSSION Topic ,

i SPES-2 Station Blackout I

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4 SPES-2 Station Blackout t

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November 4,1993 staff letter requested "A station blackout, resulting in a cooldown to ' safe shutdown' conditions followed by ADS actuation after a simulated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

Subsequent discussion with the staffindicated that the staffis interested in How the plant will respond after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the timer actuates

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Plant pressure and power would be lower IRWST will be at or near saturation IRWST may not be full 1

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W Evaluation ofIssue

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  • AP600 SSAR Chapter 15.6 included results for an inadvertent ADS actuation scenario

- SSAR case is initiated at full power RCS conditions of

" Pressure = 2251 psia

" T, = 599.3*F 3o  ;

    • T. , = 564'F i

" Core Power = 1971.7 MWt Tmm @ time ofIRWST inj. = 220*F (Bottom node = 120 F)

The automatic ADS actuation 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a station blackout is bounded in severity by the SSAR inadvertent ADS case '

i Depressurizing to and achieving IRWST injection is easier after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Decay heat at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is only 40% of SSAR value at time of 4th stage valve actuation

- After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IRWST tank level is only expected to decrease approximately 3-4 feet

" Effect on IRWST injection due to these variation in level and temperature is expected to be small  ;

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after station blackout, the CMTs water is expected to reach thermal equilibrium with RCS cold leg due to recirculation

" Injection of CMT proceeds with flashing as system depressurizes t

- Draindown of CMT is expected to proceed similar to the small LOCA (1" break) case analyzed in SSAR

" 1" break SSAR case exhibits CMT draindown with flashing for significant portion of the transient

  • SPES-2 test matrix include a small break s 1" as well as ADS actuation pre-operational tests that will be used for code validation.

i S PES-test wpf i

i W Evaluation of Issue (Cont.)

W has performed a station blackout scoping calculation followed by ADS actuation after 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> with the MAAP4 code.

The RCS conditions at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> are significantly lower than the full power conditions due to cooling by the PRHR '

Pressure = 290 psia T, = 340'F Core Power = 10.4 MWt T mwer @ inj. = 245'F i After 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> IRWST tank les el has only decreased approximately 3 feet (condensate returning ttc the IRWST is conservatively not modeled)

Stable IRWST injection predicted Impact on IRWST injection due to level and temperature variation is small l

l SPE54st wpf

PRELIMINARY .

MAAP4 SCOPING ANALYSIS AP600 24 hour Station Blackout. ADS at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> Reactor Coolant System Pressure RCS 2500 ,

7 2000 -]

~

m -

o_ -

v 1500 _

e -

u a 1000

~

\

m -

500 N x N

0 0 5 10 15 20 25 30 Time (hrs)

PRELIMINARY

  • MAAP4 SCOPING ANALYSIS AP600 24 Hour Station Blackaut. ADS at 22 Hours Average RCS Water Temperature RCS Tavg 600 _ ,

m 550  ;

u_  :

500 _

o,  :

[ 450 E" iK m

350 E _ ,

a)  :

300 250 0 5 10 15 20 25 30 Time (hrs)

PRELIMINARY .

MAAP4 SCOPING ANALYSIS AP600 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Station BIackout. ADS at 22 hours IRWST Woter Temperature IRWST m 260 _

240 f

~

G) -

u 220 _

a -

O 200 -

' 2 e 180

~ =

E -

/

a, 160 140 5/

e

- 120 _

o _

3: 100 ~ ' ' ' ' ' ' ' ' ' ' ' ' ' i '

0 5 10 15 20 25 30 Time (hrs) i

PRELIMINARY

  • MAAP4 SCOPING ANALYSIS AP600 24 hour Station Blackout, ADS at 22 hours IRWST Water Level IRWST 36 ,

m

~

34 N

, 32

\ N

> ~

G)

J 30 u

O w -

0 28 3:

26 0 5 10 15 20 25 30 Time (hrs)

4 PRELIMINARY MAAP4 SCOPING ANALYSIS -

AP600 24 hour Station Blackout. ADS at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> IRWST Injection Rate IRWST injection m 80 ,

u> -

N -

E l

n I '-

60 f 1 u

%t cr 40

~ 20 o _

G)

C -

0 O 5 10 15 20 25 30 Time (hrs)

1 1 IRWST FLOW RATE -

l i

e M

m 3e U

t B {

d

  • e ,

_ 1 l

I "g i

^ '

9 e i i s 3W N a

N e

N i

N 4

6 0 SCO 1000'

- TDC CcLCS 3 4ft90 SELOS - -

AP600 INADVERTENT ADS - SSAR RESULTS

e .,

l l

W Evaluation of Issue SPES-II Facility Consideration 4

SPES.II currently does not have capability to artificially heat the IRWST to saturation conditions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

- Facility modification will delay test schedule a

Due to facility metal heat, heat loss, and other limitations, it is not practical to run a test out for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Limited useful information is expected. l The SPES-2 test matrix does include a hot pre-operational test (H.06) which l is equivalent to an inadvertent ADS event from full power, full pressure as well as a hot pre-operational test (H 05) which actuates the ADS from low '

pressure.

These two tests will provide adequate information of the ADS actuation from different conditions.

Sufficient data available for code validation.

t S PES-ten eyf

4 ..

SPES-2 Station Blackout Summary

  • W has presented an inadvertent ADS actuation case in the SSAR which ,

bounds in severity of ADS actuation at a later time  !

Effect on IRWST injection due to variations in IRWST level and temperature is small W has performed MAAP4 scoping calculations which demonstrate successful IWHST injection for ADS actuation after -22 hours SPES-2 test matrix includes pre-operational tests that actuate ADS from full ,

and low pressure i 1

- Adequate information on ADS actuation from different conditions i available l

Limited useful information is expected from the proposed station i blackout case

)

l Recommendation W recommends that station blackout tests not be added i

s PELieu wpf j

I

,. +

a Westinghouse /NRC Test issues Meeting i

, December 10.1993 OWFN 1-F-19 (8:30 - 4:00)

1. Introduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung ATWS A. C. Cheung Westinghouse
4. OSU Tests Higher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica
5. Noncondensable gas issues L. E. Hochreiter
6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closeout Summary of Remaining Issues & Actions All

y &

TEST ISSUE DISCUSSION Topic SPES-2 ATWS I

l I

5PLC Al% 5 WPF 1

e . -;

SPFE-2 ATWS -

Ilackstround November 4,1993 staff letter requested W to simulate an

" Anticipated transient without scram (ATWS)"at SPES 2 because it is the only facility capable of scaling full power  ;

q a

The staff letter did not clarify the reasons for requesting this test or the specific new phenomena that need to be validated e

hP6$1'AT% 5 WPF

. .o.

I

.~

W Evaluation ofIssue t Current Status l

l t

Section 15.8 of AP600 SSAR delineated how AP600 complied with the ATWS rule (10 CFR 50.62)

A diverse actuation system which provides for all AMSAC protection features mandated for W plants plus a diverse reactor scram -

Risk of core damage due to ATWS is well below the 1 x '

104 /R-yr goal I

In RAI 440.26 the Staff requested W submit an ATWS analysis .)

l In response W agreed to perform a limiting ATWS analysis (a complete loss of normal feedwater (LONF) ATWS) to j demonstrate:

AP600 ATWS response characteristics are comparable to other W plants j l  !

l -

Bases considered by the Staff in formulation of 10 CFR 50.62 is applicable to AP600 I

I l

sPESL ATWS WPF I

~

W Evaluation ofIssue l Limiting ATWS Analysis

. A typicallimiting ATWS event resulting from a reduction in the heat removal capability of the secondary side has been performed.

Complete loss of normal feedwater (LONF) ATWS case is analyzed The AP600 LOFTRAN code is used '

Analysis methodology and assumptions used consistent with previous submittal (WCAP 11992, and WCAP 11993)

L Both cases with and without diverse reactor scram actuated by the DAS studied )

P (i'ES?A3 $ WPF

  • t Q t

W Evaluation of Issue l LONF ATWS Results  !

-i

. Case Not Crediting Diverse Reactor scram i Only actuation of the PRHR system and the turbine trip functions of the DAS is assumed for this case i

Plant transient response similar to other W PWRs for same event

- Sensitivity to MTC value studied P

MTC value that gives a peak RCS pressure of approximately 3200 psig is determined to be -7.3 pcm/"F Magnitude is comparable to previous submittal for other  :

W plants I

  • Case modeling diverse reactor scram l

Reactor scram initiated by DAS modeled  :

1 1

- MTC of -7.3 pcm/'F used '

I

- DAS generated reactor scram at 75.4 seconds l

Peak pressure reached is 2571 psia i 1

1

- DAS eliminates any challenge to the pressure limit l i

)

. AP600 is in compliance with the ATWS rule I 1

- The ATWS analysis will be submitted to Staffin early 1994 l l

i

l J

I

_ LONF-ATWS  !

REPRESENTATIVE W PLANT-I LONF HI TAVG WITH AMSAC ON LOW.SG LEVEL AT 19 $ ECON 05 PLOT 1 RUN 2 e 1.4 g 1.2 >

1'  :

b "

W gm  :

u r 5 '

e >

5  :

14

.2 0.

0 50 100 150 200 250 300 350 400 450 500 550 600 .

TIME (SEC) u,w,w <

i 4

a o 1

~

l LONF-ATWC  :

REPRESENTATIVE V( PLANT LORF HI TAVQ WITH AMSAC ON LOW SG LEVEL AT 19 SECONOS PLOT 4 MUN 2 3000.

2900. -

E 2800. -

t 2700. '

, 2600. -

2500.< I

=

$ 2400. -

2300. '

i r Z200. -

2100.< / ,

2000.

O. 50. 100. 150. 200. 250. 300. 350. 400. 450. 500. 550. 600. I TIME (SEC) u,usu

/

1

e e e

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1, Figure 1-1 AP600 LONF ATWS: Base Case Nuclear Power 1.2

' N __

g 0.8 -

O

=

E

$ 0.6 --

t 5

z 02 _

0.2 -

0 O 50 100 150 200 250 300 Time (sec)

PRELMINARY 440.26(RI)-11 i

I l

l

e

  • NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 i

3 L

Figure 15 AP600 LONF ATWS: Base Case Pressurizer Pressure 3,400 ,

3.200 -- - -

p 3.000 -

d e

?, 2.800 -

3 -

I

  • 2,600 -

o 2

e 1 2,400 -

2200 -

2.000 O 50 100 150 200 250 300 Time (sec)

PRELIMINARY 440.26(RI)-15

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 2-1 AP600 LONF ATWS: Diverse Scram Assumed Nuclear Power 1.2

' N g 0.8 -

O e,

5

[0.6 E

2 8

z o4 -

0.2 -

0 O 50 100 150 200 250 300

  • wne (see)  ;

a PRELIMINARY l

uo.26(R1)-19 i

c e e NRC REQUEST FOR ADDITIONAL INFORMATION Round: 0 Question Set: 09/23/92 Response Revision 1 mm .

Figure 2 5 AP600 LONF ATWS: Diverse Scram Assumed Pressurizer Pressure 3.200 --

p 3,000 e

j 2,800

. ~

$ 2,600 --

s I 2,400 --

2,200 -

2,000 O SO 100 150 200 250 300 Time (sec) o PRELIMINARY 440.26(RI)-24 W Westingh00se

4- #

l l ~

W Evaluation ofIssue SPES II Facility Considerations

= Present SPES-2 Facility has no capability to simulate the nuclear '

feedback essential to an ATWS event

= A heatup test without the proper moderator and fuel feedback will not provide valid ATWS transient data for code validation 1

i l

l I

1 sPES:iAT4 5 E PF l

  • )

[

Summary .

ll AP600 design includes a DAS that complies with the ATWS rule

. Limiting ATWS event performed  !

AP600 plant response to the event is similar to other W PWRs MTC value is comparable to other plant j J

Reactor scram initiated by DAS eliminates any challenge to pressure limit i

. SPES-2 does not have capability to simulate an ATWS event '.

- Cannot generate valid data for code validation for an ATWS event l

Recommendation i

+ T/H phenomena occurring during an ATWS in AP600 are same as in l current plants for AP600 l

- No additional data for validation of existing code and methods needed SPEC, AT4 5 WPF

e 9 5

l Westinghouse /NRC Test Issues Meeting j December 10,1993 OWFN 1.F-19 i (8:30 - 4:00) ,

1. Introduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter t
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung l Station Blackout A. C. Cheung ATWS A. C. Cheung t

l l

4. OSU Tests l

I -

Higher Risk Shutdown event T. L. Schulz Westinghouse Capability to model SGTR and MSLB E. J. Piplica ,

i 5. Noncondensable gas issues L. E. Hochreiter

6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closcout Summary of Remaining Issues & Actions All l

l l

I

OSU SHUTDOWN EVENT Q NRC Request Want an OSU test that demonstrates the capability of the safety and non-safety systems to deal with higher risk shutdown incidents

- Presentation Passive system shutdown capabilities Recovery from use of passive systems Shutdown PRA Summary / Recommendations TLS I .1 J 9, 9 3

--- - . - - - _ . = _ _ _ = - _

w, AP600 SHUTDOWN CAPABILITIES .

'L. J Passive Safety Functions Provided During All Shutdown Modes

- Hot Shutdown / Hot Standby / Cold Shutdown (SSAR sec 6.3.3.4.1)

Protection provided by PRHR HX, CMT, IRWST, and ADS Required by Tech Spec

- Cold Shutdown Reduced Inventory (SSAR sec 6.3.3.4.3)

PRHR HX ineffective (RCS open)

CMT / accum unnecessary (RCS atm pres)

Protection provided by:

IRWST injection (MOV closed, operable)

ADS stages 1,2,3 open Containment integrity Required by Tech Spec

- Refueling Shutdown (SSAR sec 6.3.3.4.4)

Refueling cavity provides >120 hours with equipment hatch open Refueling cavity takes >9 hours to boil TIE 12 / '3 / 9 3

por* '

SHUTDOWN RECOVERY L.

Recovery From Use of Passive Systems During Shutdown Event

- Recovery From Reduced inventory Event Passive systems providing protection IRWST injection, ADS, passive cont cooling Refill RCS from IRWST using RNS (if required)

Simple 3 valve alignment Cool RCS using RNS to recirculate RCS Simple 3 valve alignment Cool containment using fan coolers TLE;12 / 9 / 9 3

pr,, .

SHUTDOWN PRA i' Shutdown Conditions Evaluated (PRA sec 7.6)

Cooldown to cold shutdown Cold shutdown with reduced RCS inventory Filling refueling cavity Draining refueling cavity Filling RCS Heatup to hot standby

- Shutdown Conditions Modeled in PRA Shutdown with RCS intact Shutdown with RCS inventory reduced TLS12/9/93

e

! !fkd '

SHUTDOWN PRA i 1i t - Shutdown Events Evaluated (PRA sec 7.6) '

Reactivity accidents (boron dilution, rod withdrawal)

Loss of normal decay heat removal Loss of offsite power Loss of reactor coolant (pipe break, inadvertent drains)

- Shutdown Events Modeled in PRA Loss of normal decay heat removal Loss of offsite power Loss of reactor coolant (pipe break, inadvertent drains) l TLS12/9/93

j!!!ilii}ili ~

SHUTDOWN PRA .

i '

Core Damage Frequency Total (PRA sec 8.4) 8.9E-8/yr (shutdown) vs 3.3E-7/yr (at power)

- Core Damage Frequency vs Event (PRA sec 8.4) 94.5% Loss power or heat removal with reduced inventory 3.2% Loss power or heat removal with RCS intact 1.1 % LOCA / drain with reduced inventory 0.8% LOCA / drain with RCS intact

- Shutdown PRA Insight Shutdown risk less than at power

- Dominant shutdown risk due to loss power / heat removal with reduced RCS inventory RTNSS identified RNS as important (WCAP-13856)

Not important to focused PRA Important to shutdown IE frequency during mid-loop Controls required on availability of RNS during mid-loop TI.S 12 9 / 9 i

k kk

~

OSU SHUTDOWN

SUMMARY

i. *.

- Passive Systems Provide Protection During All Shutdown Mddes

- RCS Drained Condition is Only PRA important Condition Passive system protection covered by Tech Spec ADS open, IRWST injection, containment integrity T/H analysis bounded by inadvertent ADS from full power Passive system operation Recovery of non-safety systems

- Westinghouse Position -- No Shutdown Test at OSU Higher risk shutdown events benign from T/H phenomena standpoint Passive system operation bounded by at power OSU tests Interaction with RNS bounded by at power OSU tests TLS12/9/93

., 'e ,

Westinghouse /NRC Test Issues Meeting December 10,1993 OWFN 1-F-19 (8:30.- 4:00) 1

1. Introduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung l Station Blackout A. C. Cheung ATWS A. C. Cheung i

)

4. OSU Tests 4 l

1 Higher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica ,

Westinghouse  !

I l

5. Noncondensable gas issues L. E. Hochreiter !
6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closecut Summary of Remaining Issues & Actions All l

l l

m,

- OSU Capabilities for Performing SGTR & MSLB Tests a t

issue: The capability should be provided to. perform SGTR & MSLB tests at the O$U facility in the event the SPES-2 comparable test results indicate a need.

4

4 OSU Capabilities for Performing SGTR & MSLB Tests c The SPES-2 facility is the preferred facility for performing SGTR & MSLB tests:

o SPES-2 is a full pressure system test - more appropriate for these transients.

o Analysis indicates that ADS will not be initiated by either a SGTR or a MSLB.

i The SPES-2 facility has been scaled and configured to perform SGTR & MSLB tests:

o SGTR & MSLB systems & instrumentation are designed and available.

o- Only SGTR hardware needs to be installed in SPES-2.

o MSLB systems are in place; successfully tested in SPES-1.

_ _ _ _ _ __--__. _______ _ _- .-.___-- _ _ _ _ _ ______----_____ _ _ - . - - _ _ - - _ __-..._._._______.m. - _ _ . _ _

OSU Capabilities for Performing SGTR & MSLB Tests e, The OSU facility has been designed to perform SBLOCA transients and to investigate low pressure, long term cooling events: ,

o OSU maximum primary side operating pressure is 400 psia at 440*F.

o OSU secondary side pressure is 350 psia c.: 430 F.

The capability to simulate a SGTR or MSLB will require additional design and facility modifications at OSU.

O OSU Capabilities for Performing SGTR & MSLB Tests '

The OSU facility could be configured to perform SGTR tests:

i o One steam generator at OSU has been fabricated to allow secondary side fluid to enter the primary side system by providing piping connections from inside the steam generator and the steam generator channel head.

o Scaling and design of the systems would have to be performed.

o Piping, valves, instrumentation would need to be installed.

o The control panel and data aquisition would need to be expanded, programmed and validated.

o The testing schedule would be extended.

O OSU Capabilities for Performing SGTR & MSLB Tests e The OSU facility could be configured to perform MSLB tests:

o Seconday side is discharged to atmosphere.

o Tests could be performed by allowing the steam generator feedwater level to fall, thereby simulating a MSLB.

o Scaling and design of the systems would have to be performed.

o Piping, valves, instrumentation would need to be installed.

o The feedwater logic would need to be reprogrammed accordingly.

o The control panel and data aquisition would need to be expanded, programmed and validated.

o The test schedule would be extended.

OSU Capabilities for Performing SGTR & MSLB Tests c' i

Westinghouse position:

i o The high pressure SPES-2 tests will provide the data needed to validate computer codes for both SGTR and MSLB transients; the facility hardware and _ ,

i controls necessary to perform the test are or will be installed.

i n

j

- _ - _ - _ - - _ - _ - - - - - _ - - _ - _ . _ - . - - . . . - - - - . ~ . . - . . _ - - - - _ _ _ . - - . . _ . - .-- =.-....:

. I Westinghouse /NRC Test Issues Meeting ,

December 10.1993 OWFN 1-F-19 (8:30 - 4:00)

1. Introduction B. A. McInt>Te
2. Core Makeup Tank Tests L. E. Hochreiter ,
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung ATWS A. C. Cheung
4. OSU Tests Higher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica
5. Noncondensable gas issues L. E. Hochreiter Westinghouse
6. PRHR tests L. E. Hochreiter
7. Check Valve Tests T. L. Schulz
8. Meeting Closecut Summary of Remaining Issues & Actions All

N S OREGON STATE UNIVERSITY LOW PRESSURE INTEGRAL SYSTEMS TEST - SPES INTEGRAL SYSTEMS TESTS - NON-CONDENSIBLE ISSUE TEST MATRIX:

"FOR THE SAME REASONS, THE FOLLOWING ACCIDENTS SHOULD BE SIMULATED WITHIN THE OSU MATRIX: 1)

CAPABILITY OF THE SAFETY AND NON-SAFETY SYSTEMS TO DEAL WITH HIGHER RISK SHUTDOWN INCIDENTS; 2) THE EFFECTS OF HYDROGEN COLLECTED ABOVE THE PRESSURIZER AND NITROGEN IN THE ACCUMULATOR ON THE '

PROGRESS OF A NUMBER OF ACCIDENT SCENARIOS; 3) TO ALLOW FOR THE POSSIBILITY THAT SPES SHOWS THAT STEAM ,

GENERATOR TUBE RUPTURE (SGTR) OR MAIN STEAMLINE I BREAK (MSLB) ACCIDENTS CAN LEAD TO ADS ACTUATION, THE .

CAPABILITY OF PERFORMING SUCH TESTS SHOULD BE i PROVIDED". l

" ITEM 2, $FFECTS OF HYDROGEN COLLECTED ABOVE PRESSURIZER AND NITROGEN IN THE ACCUMULATOR ON THE PROGRESS OF A NUMBER OF ACCIDENT SCENARIOS".

I NOTE SIMILAR OUESTION ON NON-CONDENSIBLES FOR SPES 0672LH 120093

F g g NON-CONDENSIBLES IN THE PRESSURIZER  ;

i i

THE CMT TESTS HAVE BF.EN SPECIFICALLY DESIGNED TO PARAMETRICALLY ADDRESS THE IMPACT OF NON-CONDENSIBLES ON CMT CONDENSATION HEAT TRANSFER i

l THE CMT TESTS WILL ALSO EXAMINE, PARAMETRICALLY THE ' i IMPACT OF NON-CONDENSIBLES ON THE RECIRCULATING BEHAVIOR OF THE CMT COLD LEG BALANCE LINE l

mnu m mas

r 1 o o NON-CONDENSIBLE EFFECTS OF THE NITROGEN IN THE ACCUMULATOR ,

i l

l BOTH SPES AND OSU WILL SIMULATE THE NON-CONDENSIBLE l NITROGEN GAS DISCHARGE FROM THE ACCUMULATORS INTO THE RCS l

THE RESULTING BEHAVIOR OF THE SYSTEM RESPONSE WILL BE OBSERVED FOR THE NITROGEN INJECTION I

0672LH.120893 l

(-

o .A  :

l t

i ISSUES OF NON-CONDENSIBLES ARE BEING ADDRESSED IN THE CMT '

TESTS FOR THE PRESSURIZER HYDROGEN EFFECTS l

THE OSU AND SPES SYSTEM TESTS WILL SIMULATE THE NITROGEN i EFFECTS, WHICH BY FAR, INJECT THE MOST NON-CONDENSIBLE GAS R INTO THE RCS, AND MAY IMPACT DEPRESSURIZATION .

J

,1 0572LH-120893 l

I

Westinghouse /NRC Test Issues Meeting December 10,1993 OWFN 1-F-19 (8:30 - 4:00)

1. Intmduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests MtJtiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung ATWS A. C. Cheung
4. OSU Tests liigher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica
5. Noncondensable gas issues L. E. Hochreiter
6. PRHR tests L. E. Hochreiter Westinghouse
7. Check Valve Tests T. L. Schulz
8. Meeting Closeout Summary of Remaining Issues & Actions All l

5 PRHR ISSUE FROM NOVEMBER 14TH LETTER PASSIVE RESIDUAL HEAT REMOVAL (PRHR) HEAT EXCHANGER PERFORMANCE DATA:

"THE C-SHAPE OF THE HEAT EXCHANGER TUBES AND THEIR ARRANGEMENT IN THE IN-CONTAINMENT REFUELING WATER STORAGE TANK (IRWST) DIFFER MARKEDLY FROM THE 3-VERTICAL TUBE TEST USED TO DEVELOP THE DESCRIPTION OF THE PRHR PERFORMANCE. JUSTIFICATION IS NEEDED TO SUPPORT WESTINGHOUSE'S CLAIM THAT NO MORE TESTING IS REQUIRED, ESPECIALLY CONSIDERING POOL-SIDE THERMO-HYDRAULICS".

i omLM2 so

. k OUTLINE DISCUSSION OF PRHR TEST RESULTS BASIS FOR TUBE SPACING l

DISCUSSION OF C-TUBE CONFIGURATION DISCUSSION ON TUBE ORIENTATION DISCUSSION OF CONCERN ON VAPOR BLANKETING ON POOL SIDE 0 6 120893

PASSIVE RHR HEAT EXCHANGER TEST p TEST OBJECTIVES a

Demonstrate / Optimize Heat Transfer Capability of the PRHR Heat Exchanger at Prototypic Tube Side (Primary Side) Flow Rates, Inlet Temperatures and IRWST Conditions (Secondary Side)

=

Assess PRHR Heat Exchanger Design Features which Optimize Steam Quenching and Tank Mixing Water Level Above Tubes Distance Between Tubes and IRWST The Use of Baffles in the Tank December 10,1993 AP600 NRC Meeting on Test issues IK6 /P 6

PASSIVE RHR HEAT EXCHANGER TEST y PRIMARY SIDE: Passive RHR Heat Exchanger o Full System Temperature and Pressure o Three Vertical Heat Exchanger Tubes; Two Tubes Fully Instrumented >

Tube Sido Axial Fluid Temperatures Tube Outside Wall Temperatures o Full Range of PRHR Tube Flow Rates o Tube Flow Rates Individually Controlled and Measured o Tube inlet and Outlet Temperatures Measured o 350 KW Electric Heater Used: Power Measured December 10,1993 AP600 NRC Meeting on Test Issues

PASSIVE RHR HEAT EXCHANGER TEST e Passive RHR Heat Exchanger Test Loop

-- V --I _ _

n

[5 y p a

_J h,1:-_

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December 10,1993 600 E Meedng on Ted bwes

PASSIVE RHR HEAT EXCHANGER TEST  ;

Tank Thermocouple Locations E

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e E M]TE e - T/C LOCATED EVERY 1.5' E - T/C LDCATED EVERY 3' December 10,1993 AP600 NRC Meeting on Test issues 0667Pe t I

_ _ . . _ _ . ._.__________________.._________________m_________ _ - . , . . - . . - _ , - - , . . _ , . . . - , . . . . - -- . ~ , - - . ,

PASSIVE RHR HEAT EXCHANGER TEST <.

Thermocouple Vertical Locations r

( f p C367 Pit?

i TUBE SPACING DIFFERENT EXPERIMENTS WERE PERFC~lMED TO EXAMINE TUBE SPACING DETAILED FLUID TEMPERATURE MEASUREMENTS WERE MADE TO DETERMINE THE " ZONE OF INFLUENCE" AROUND A SINGLE TUBE EXPERIMENTS WERE PERFORMED WITH 1,2,3 TUBES OPERATING A BAFFLE WAS PLACED IN FRONT OF AND BEHIND THE TUBES TO BLOCK CROSSFLOW INTO THE HEATED BUNDLE wwnn

P ,? g TUBE SPACING STUDIES - RESULTS:

ZONE OF INFLUENCE WAS FOUND TO BE - Pld = 1.5 - 2.0 -

A VALUE OF Pld = 2.0 WAS CHOSEN FOR THE DESIGN wru.umn

AP800 PRHR TEST - Plume Test P-02 ,4

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TOP .* MIDDLE x BOTTOM

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AP600 PRHR TEST - Plume Test C-04 .

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-=- TOP + MIDDLE = BOTTOM

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f EXPERIMENTS WITH 1,2,3 TUBES OPERATING -- RESULTS .

HEAT FLUX BEHAVIOR WAS THE SAME WITHIN THE DATA SCATTER FOR 1,2, OR 3 TUBES OPERATING INDICATING THAT THEY PERFORM INDEPENDENTLY 1

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0672LH-120893

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w Figure 7.3-7:- Local Wall Heat Flux for Steady State Test S-05 (3 Tubes @ 1 gpm each) oo609 3;ionis2991 7-25  ;.

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'4 EXPERIMENTS WITH DIFFERENT BAFFLE LOCATION - RESULTS AS THE BAFFLES WERE PLACED CLOSER TO THE TUBES, TO SIMULATE THE MAXIMUM BLOCKAGE EFFECT OF OTHER R0WS OF TUBES, THE HEAT TRANSFER WAS SLIGHTLY ENHANCED DUE TO THE INCREASED PUMPING ACTION OF THE HEATED TUBES 0672LH-120093 i

r 3 PRHR HEAT EXCHANGER CONFIGURATION TESTS g'f '

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TUBE ORIEhTATION EFFECTS A PORTION OF THE C-TUBE HEAT EXCHANGER IS HORIZONTAL DATA IN THE LITERATURE INDICATED THAT HORIZONTAL TUBES HAVE HIGHER HEAT TRANSFER THAN VERTICAL THE EXCHANGER AREA IS - 56.5% VERTICAL, WITH 21.7%

HORIZONTAL AT FACH END.

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0572L4120093

I Figure 15 CONVECTION AND BOILING HEAT TRANSFER FROM WIRES AT DIFFERENT ANGLES (REFERENCE 1) 70 - /,

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PASSIVE RHR HEAT EXCHANGER TEST _

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PASSIVE RHR HEAT EXCHANGER TEST <

PRHR HX Tube Spacing in Horizontal Bundle d j w' r1.5" 1.5 -

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Plan View Section View December 10,1993 AP600 NRC Meeting on Test Issues

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PASSIVE RHR HEAT EXCHANGER TEST ,

PRHR HX Tube Spacing in Horizontal Bundle d s 1.5" ,

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  • eo993 resoo nne ueeting en rest issues g

SECONDARY SIDE VAPOR BLANKETING PEAK HEAT FLUXES (VAPOR GENERATION) OCCUR AT PUMPED (HIGH FLOW) CONDITIONS AT THE HIGHEST INLET TEMPERATURES, l.E., WHEN ON-SITE POWER IS AVAILABLE FOR NATURAL CIRCULATION CONDITIONS, THE PEAK HEAT FLUXES (VAPOR GENERATION) IS SIGNIFICANTLY LOWER IN EITHER CASE, THERE !S AMPLE MARGIN TO CHF AND NONE WAS DETECTED IN THE TESTS OG72LH-120893

SECONDARY SIDE VAPOR BLANKETING ,

SINCE THE PRHR HAS A MORE OPEN TUBE LATTICE (P/d = 2, Pld = 4)

THE HEAT FLUXES ARE LOW, PARTICULARLY FOR NATURAL CIRCULATION CONDITIONS ON THE PRIMARY SYSTEM WE DO NOT BELIEVE THAT VAPOR BLANKETING OF THE TUBES WILL  !

OCCUR j i

0672LK120093 l

WESTINGHOUSE POSITION VAPOR BLANKETING NOT EXPECTED TO BE AN ISSUE FOR PRHR ADDITIONAL INVESTIGATION OF OPEN LITERATURE WILL BE PERFORMED TO PROVIDE ADDITIONAL CONFIRMATION NO ADDITIONAL TESTING NECESSARY 0672LH-120093

Westinghouse /NRC Test Issues Meeting December 10.1993 OWFN 1-F-19 (8:30 - 4:00)

1. Introduction B. A. McIntyre
2. Core Makeup Tank Tests L. E. Hochreiter
3. SPES-2 Tests Multiple SGTR with ADS A. C. Cheung Station Blackout A. C. Cheung ATWS A. C. Cheur g 4 OSU Tests '

Iligher Risk Shutdown event T. L. Schulz Capability to model SGTR and MSLB E. J. Piplica

5. Noncondensable gas issues L. E. Hochreiter
6. PRHR tests L. E. Hochreiter l
7. Check Valve Tests T. L. Schulz Westinghouse
8. Meeting Closecut Summary of Remaining Issues & Actions All s

CH ECK VA_VE TEST \lG E' NRC Request '

Performance of check valves is critical to the CMTs, accumulators, and the IRWST. The staff must have assurance that the IRWST check valves will open under low DP following long periods of operating closed under a larger DP. Want plans for performance and reliability qualification testing. Also ISI/IST procedures that are necessary to maintain qualification throughout plant life.

- Presentation AP600 use of check valves Check valve reliability Planned check valve tests Summary / Recommendations TLE. l ' G G1

a CHEC< VALVE USE E Check Valve Well Suited to Passive Systems '

Self actuating (passive)

Simple (no support systems required, I&C/ power)

Reduced ISI/IST requirements

- Similar Check Valves Used in Current Plants Simple swing disk design Stainless steel body with stellite seats RCS water chemistry Standby operation Design improvements Position indication Avoidance of improper assembly Inservice testing TLS12/9/93

CHECK VALVE APPLICATIONS O

- Accumulator Discharge Swing disk, normally closed >

1550 psi close DP, significant opening DP Inservice test each refueling Prototypical opening DP, full open flow

- CMT Discharge Tilt disk, normally open No DP normally Inservice test each refueling Prototypical forward flow Prototypical closing flow IRWST Injection Swing disk, normally closed 2250 psi close DP, low opening DP Inservice test each refueling Prototypical opening DP, flow TLS12/9/93 L . _ . . _ _ . _ _ . ._ _ _ _ __J -

CHECK VALVE RELIABILITY []

Westinghouse / Penn State Reliability Evaluation <

NPRDS records show 4500 check valve failures 1984 to 1990 Only 87 of these were failures to open None of these failures was for a simple stainless check valve in standby service No indication of boric acid corrosion or self welding NRC Check Valve Reliability Evaluation (NUREG/CR-5944)

Same database (NPRDS)

Valve age not important 7% of failures stuck closed; most in CS, small, vacuum breakers Did not provide multiple variable sorts Similar trends with Westinghouse evaluation

- Check Valves Are Very Reliable in Such Service TLS12/9/93 L -_ _- -__-__ -______ _-__ - _ -___ _.-_- -___ -__ - _ - _ - . . - . . a

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CHECK VALVE TESTING PLA\l [-1 Design Certification Testing >

Low DP performance completed New valves,3 in series (IRWST/recirc)

Flow range (none to full open)

Opening DP, disk behavior Low DP opening of aged valves, in-plant 2 valves tested Others planned Equipment Qualification Testing Extensive vendor specific AP600 in-Plant Testing Plant startup l

Prototypical flow / DP Plant inservice, each refueling Prototypical flow / DP TLs12:9293 L__--__-________--__-___ _ _ _ _. _ _ _ _ _. _--._-____-__.-J

N 1

CHECK VALVE PRA RELIABILITY O i

- Check Valve Reliability Used in AP600 PRA <

Accum discharge,1 failure in 570 demands Based on EPRI KAG with 2 yr test frequency IRWST injection,1 failure in 115 demands Based on EPRI KAG with 2 yr test frequency (severe duty)

Check Valve Reliability Will Be Given To Valve Vendor In Equipment Spec Not requirement

- Check Valve Reliability Used in PRA Supported by design process Supported by testing Design certification Equipment qualification AP600 startup AP600 inservice TLS12/9/93

q CHECK VALVE IN-PLANT TESTING O#

First Test (Farley) >

2 - 6" high pressure swing disk check valves in HPSI High opening DP due to improper test Confirmed by re-test Re-test with proper conditions showed low opening DP

- Other Tests in Planning TLS12/9/93 L_. _

m y

y CHECK VALVE

SUMMARY

pre!

Check Valves Will Be Reliable in AP600 >

Through design / specification process Proper check valve application Through test program Design certification, equipment qualification, and AP600 plant startup and inservice tests Check valve reliabilities used in PRA are appropriate Westinghouse Position Test of several more aged check valves (in plants) under design certification Describe IST for these check valves (RAI 210.24)

No additional testing required under design certification i

i TLs12/9/93

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. . . . . _ . - _ .

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