Text

Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 from Ltrs of 931027,1102 & 30.Listing of NRC Requests for Addl Info Responded to in Ltr Contained in Attachment a
	ML20063D492
Person / Time
Site:	05200003
Issue date:	02/03/1994
From:	Liparulo N; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Borchardt R; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NTD-NRC-94-4059, NUDOCS 9402080181
	Download: ML20063D492 (115)
	v • d • e

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Westinghouse Energy Systems Ba 355 Pit: sbvgh Pemsylvama 15?30-0355 '

Electric Corporation NTD-NRC-94-4059 DCP/NRC0001 Docket No.: STN-52-003 February 3,1994 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 A'ITENTION: R.W.BORCHARDT

SUBJECT:

WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600 ;

Dear Mr. Borchardt:

Enclosed are three copics of the Westinghouse responses to NRC requests for additional information on the AP600 from your letters of October 27,1993, November 2,1993 and November 30,1993.

This transmittal completes the responses to these letters. In addition, revised responses for a number of previously provided responses are included.

A listing of the NRC requests for additional information responded to in this letter is contained in Attachment A. Attachment B is a complete listing of the questions associated with the November 2, 1993 and November 30,1993 letters and the corresponding letters that provided our response.

These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Hasselberg's copy. >

If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.

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0Ylkh Nicholas J. Liparuto, Manager l Nuclear Safety & Regulatory Activities

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-NTD-NRC-94-4059 ATTACHMENT A AP600 RAI RESPONSES SUBMITTED FEBRUARY 3,1994 RAI No. Issue 250.015R01i Remote inspection features 260.011 : Applicable revision to WCAP-8370 ,

260.012 ; Quality assurance on D-RAP systems ;

. 260.013 : QA on SSCs not required to meet Appendix B 260.014 : Commitment to meet GL 85-06 260.015 QA program for fire protection . ,

260.016 QA requirement on Class D/seisraic Cat I _,

260.017 : Requirements for construction 260.018 ; QA applied to software in the design phase 260.019 : QA applied to test programs 410.023R01: First stage. ADS Hydrostatic Loads 440.013R01; ADS testing 440.036 ; Purpose of val res V007A,-B,-C and -D 440.037 : PRHR heat exchanger heat transfer 440.038 : Hydrogen venting from the pressurizer 440.039 : " failure of a CMT discharge valve" -;

440.040 : MSLB analysis, feed flow assumptions 440.041 : MSLB minimum DNBR plot- d 440.042 ;- Feedline break PRHR delay rationale 440.043 : Locked rotor failed fuel assumptions 440.044 : Rationale for locked rotor being limiting 440.045- t Rod ejection failed fuel assumptio'ns 440.046 I Inadvertant 4th stage opening at pressure 440.047 1 ' SGTR-SG overfill analytical results -

440.048 l Assumptions on available equipment / boron dilution 470.009R01i LOCA doses using NUREG-1465 assumptions ,

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NTD-NRC-94-4059 ATTACIIMENT A AP600 RAI RESPONSES SUBMITTED FEBRUARY 3,1994 RAI No. Issue 471.021 : Update radiation zone maps 480.038 : Igniter placement criteria 480.039 : Igniter coverage of hydrogen source locations 480.040 : Subcompartments without igniters 480.041 l Criteria on igniter separation 480.042 : Igniter electrical power source 480.043 : Structural response / detonation calculations 480.044 : Potential for large hydrogen concentration 480.045 : Inniter reliability given one per room 480.047 : Review of hydrogen experimental databases 480.048 : Eauipment survivability in " burn zones" 620.045R0l! Review of operatina conditions 630.003R0l! D-RAP Desian Organization

630.004R01: Priority of safety goals 720.056R02: Accident management issues 720.070R01: Use of PRA insights in ITAACs, DAC. and DRAP 952.012 : Addition of flowmeter to ADS Phase B design 952.013 : Effect of non-condensible cases in test prograns 952.014 ! Applicability of PRIIR test to current design 952.015 : ADS Phase B capability to hold constant pressure 952.016 : SPES-2 design documents update -

952.017 : Insights from pre-operational SPES-2 tests 952.018 : Cold and hot pre-operational data from SPES-2 952.019 : Flow orifices at SPES-2 952.020 : Volume versus elevation for 3 PES-2 952.021 : SGTR analysis for hot and cold side breaks 2

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NTD-NRC-94-4059 ATTACHMENT A AP600 RAI RESPONSES SUBMITTED FEBRUARY 3,1994 RAI No. Issue 952.022 : Information nained from ADS Phase A tests ,

952.023 : Effect ofPRHR heat exchanger on ADS operation 952.024 : Test procedures for ADS Phase B testina .

952.025 : Analyses for ADS Phases A and B 952.026 : Fluid conditions for ADS Phase B tests 952.027 : RCS behavior versus ADS test facility 952.028 : Dynamic reflection wave forces in ADS operation 952.029 : SPES-2 bend / elbow radii 952.030 : SPES-2 valve tvoes and sizes 952.031 : SPES-2 schematics and drawings update -

952.032 : SPES-2 operational parame:ers 952.033 : SPES-2 component position relationships 952.034 : SPES-2 test procedures 952.035 : SPES-2 facility insulation information [

952.036 : SPES-2 annular downcomer b

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Pnnted: 02/0194 ATTACHMENT B CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTERS OF NOVEMBER 2,1993 AND NOVEMBER 30,1993 Queston issue NRC Westmghouse No. Letter Transmrtial Date 260 011 AppkcaNo revWon to WCAP-8370 11/0293 020194 260 012 Quality assurance on D RAP systems 11/0293 020194 260.013 QA on SSCs not requred to meet Appendtx B 11/0293 0203S4 260 014 Commitment to meet GL 85-06 11/02S3 0203S4 260.015 QA program for fra protectm 110293 02/0194 200 016 QA requrement on Cisne D!seestre Cet i 11/0293 020194 260 017 Requrements for eenstructon 11/02S3 02/0194 ,

260 018 QA apphed to software in the design phase 11/0293 02/03S4 260.019 QA applied to test programs 110293 0203S4 440.036 Purpose of vatves V007A,-B,4 and -D 110293 020194 440 037 PRHR heat exchanger host transfer 11/02/93 0203G4 440 038 Hydrogen venhng from the pressurtzer 11/02/93 020194 440 039 "fe4ure of a CMT decharge vatw* 1102S3 02/0194 440 040 MSLB analysas, feed flow soeumpoons 11/02S3 020194 440.041 MSLB mrwnum DNBR pkr 11/02/93 020194 440 042 Feedkne break PRHR deiey rationaio 11/0293 020194 440.043 i N rotor failed fuel assumphons 11/02S3 020194 440 044 Ratiormie for locked rotor beeng hmthng 11/02/93 0203S4 440.045 Rod ejection failed fusi assumptions 11/02/93 0203S4 440 046 Inadvertant 4th stage openrig et pressure 11/02 S3 02/0194 440 047 SGTR-SG overfall anatytcal results 11/02/93 02/0194 440 048 Assumptons on symitable equipmerttoron dilubon 11/02/93 02/0194 480 038 Igniter piecement cdteria 1142S3 020194 480.039 Igniter coverage of hydrogen source locatons 110293 020194 480 040 ShMi towithoutignitors 1102/93 OWO3S4 483 041 Critene on igniter separabon 110293 02/0194 480.042 Igniter electrical power source 11/0293 0203S4 480 043 Structural responee/detonobon calculeuans 11/02 S3 02C3S4 480.044 Potentini for large hydrogen concentraton 11/0293 02C194 480 045 Igniter rehatsty grven one per room 11/02S3 0203S4 480.046 Igniter in Techncal Specificehons 11/02/93 01/12 S4 480 047 Review of hydrogen expenmental databases 11/02S3 02/0194 480.048 Equipmerd survhrability in " bum zones" 110293 02/0194 952.012 Additen of flowmeter to ADS Phase B design 1102/93 0203S4 952 013 Effed of norH:endensable genee in test program 11/02 S 3 020394 952 014 AppucatAty of PRHR test to current design 11/02/93 0203S4 952.015 ADS Phase B capabinty to hold constant pressure 11/02/93 020194 952 016 SPES-2 design documents update 11/0293 020194 952.017 insights from pre-opershonal SPES-2 tests 11/02/93 02/03 S4 952.018 Cold and hot pre-operatonal data from SPES-2 11/0293 02/03S4 952.019 Flow artfices at SPES-2 1102S3 020194 952.020 Volume versus eievehon for SPES-2 11/02S3 0203G4 952.021 SGTR anetynas for hot and cold side treeks 11/02/93 02/0194 952 022 Informehon geened from ADS Phase A tests 110293 0203S4 952.023 Effect of PRHR heat exchanger on ADS operation 11/02/93 72/0194 952.024 Test procedures for ADS Phase B testing 1102S3 12/0194 952.025 Analyses for ADS Phases A and B 11/0293 C&294 952.026 Fkald conditions for ADS Phone B tests 11/02S 3 02/03 S 4 952.027 RCS behavior versus ADS teet faciitty 11/02/93 020194 952.028 Dynome reflechan www forces L .3S opershon 1102S3 020194 952.029 SPES-2 bend / elbow radit 11/3043 02/0194 952.030 SPES-2 weeve types and sizes 11/3G93 020194 952.031 SPES-2 ochemetes and drawings update 11/30 S 3 02/0194 952 032 SPES-2 opershonal parameters 11/3193 02/03 S4 952.033 SPES-2 component pooltion reisuonships 11/3093 020194 952.034 SPES-2 test procedures 11/30S3 020194 952 035 SPES-2 facilary insulebon informaton 11/30 S3 020194 952 036 SPES-2 annular downcomer 11/3093 02C3S4 Records pnnted 58 Page 1

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NRC REQUEST FOR ADDITIONAL INFORMATION Eu na ;

Response Revision 1 Question 250.15 Describe the features incorporated in the design that enhance inspection of the steam generator tubes without manned :

entry. Discuss whether the design features support the use of current robotic equipment used in steam generator !

tube inspection and repair. In addition, discuss whether verification have been performed, by computer simulation ,

and/or mockup, to ensure that the design will facilitate not only the use of robotic manipulators in inspecting all of the tubes within the steam generator but also in inserting the robotics into the steam generator (Section 5.4.2).

Response

The cylindrical portion of the channel head just below the tubesheet facilitates the use of robotically delivered ;

inspection and repair tooling to tube locations on the periphery of the tube bundle. The ability to reach all locations !

has been verified using computer simulations. De channel head and primary inlet and outlet nozzles have i provisions to facilitate the robotic installation of nozzle dams. The use of 18 inch diameter manway openings i provides that any equipment that is used in operating steam generators can be used in the AP600 steam generator, j

- The fourth paragraph of SSAR Subsection 5.4.2.5 will be revised to reflect this response as follows: l l

I SSAH Revision:

The steam generators := Q!p:d : permit access to tubes for inspection; smikae-repairi or plugging, if L f necessary, per the guidelines described in Regulatory Guide 1.83. He AP600 steam generator includes features :

to enhance robotics inspection of steam generator tubes without manned entry of the channel head. These include Li

- a cylindrical section of the channel headiprimary manways, and provisions lt6 facilitate the; remote installation of 'l nozzle dams. Computer simulation using' designs, of existing robotically delivemd inspex. tion and' maintenance

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j equipment. verifies that tubes can.be' accessed.

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250.15R1-1 W-Westinghouse .

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e-NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.11 Section 17.3 of the SSAR states that Revision 12 to WCAP-8370, " Energy Systems Business Unit - Power Generatmn Bustress Urut Quality Assurance Plan," will apply to future design, procurement, fabncation, inspection, and testmg activities. 'Ihis revision, now labelled Revision 12A, dated Apnl 1992, has been accepted by the NRC by letter dated Aprd 23,1992. Clarify which revision to WCAP-8370 is applicable to the AP600.

Response

Section 17.3 currently states, " Revision 12 . has been accepted by the Nuclear Regulatory Commission. Upon implementation at a date to be determined, this program will be designated as Revision 12A arxl will apply to future design, procurement, fabrication, inspection, arxl tesung activities.'

Revision 12 A to WCAP-8370 was implemented effective November 30,1992 per letter NRC-92-3708 (N. J. Liparulo to G. G. Zech, dated June 15, 1992) and has applied to AP600 work smce that time.

Revision 1 of the SSAR states that Revision 12A was implemented effective November 30,1992.

SSAR Revision: NONE 260.11-1 W Westinghouse I

NRC REQUEST FOR ADDITIONAL INFORMATION

- :in Ouestion 260.12 Table 16.2-1 of the SSAR describes timse D-RAP non-safety-related systems that provide defense-in-depth or that are used m the PRA evaluation to provide credit for event mitigation. Table 3.2.3 shows, by AP600 class, those systems for which Apperrlix B of 10 CFR Pan 50 applies including most of the D. RAP systems. Desenbe the quality assurance requirements for those D-RAP systems of the AP600 that are not covered by Table 3.2-3.

Response

SSAR Revision 1, dated January 1994, includes revisions to Table 16.2-1 to retlect the implementation of the regulatory treatment of nonsafety-related systems process. The RTNSS process implementation ts documented in WCAP 13856. ' Die nonsafety-related systems identi6ed by the RTNSS process as important are included in Table 16.2-1 and m the D-RAP. These nonsafety-related systems are classified as AP600 Class D. This classification corresponds to Quality Group D of Regulatory Guide 1.26. The quality standards identified by the regulatory guide as appbcable to Quality Group D SSC are applied to those AP600 SSCs classified as AP600 Class D except AP600 applies API 610 or Hyciraulic Institute Standards to pumps instead of " manufacturers standards" as reconunended by Regulatory Guide 1.26. In addition, AP600 applies NEMS MGl 1972 (National Electric Manufacturers Association) to ac motors and generators, and ANSI /IEEE C37,1989 to circuit breakers, switchgear, relays, substations, and fuses. These items are not addressed in Regulatory Guide 1.26.

Quahty assurance requirements for the diverse actuation system are discussed in the response to RAI 260.14.

SSAR Revision: NONE 2mN W- WeStinEhouse

l NRC REQUEST FOR ADDITIONAL INFORMATION h i

.es Quesjon 260.13 Note (2) for Table 3.2-1 of the SSAR states that the NNS defined in the ANSI 51.1 standard is divided into two AP600 equipment classifications; namely, Class D and Class NNS. It is also noted in Table 3.2-1 that Appendix B to 10 CFR Part 50 does not apply to Class D. However, Class D contains items imponant to safety for which a quality program should be implemented that will le sufficient to ensure that the functions of the equipment sinll be achieved. Also, Criteria II of Appendix B to 10 CFR Pan 50 states that the quality assurance program shall provide control over activities affecting the quahty of the identified structmes, systems, and components (SSCs) to an extent consistent with their importance to safety. Describe, in general, the quality assurance requirements for SSCs of the AP600 that are not required to meet Appendix B of 10 CFR Pan 50, but that are important to safety, such as noted in Q26012 above or in AP600 Gass D (See also Q260.14 and Q260.15 below).

Response

As indicated in SSAR Appendix 1 A, important, nonsafety-related systems and components that function as a first hne of defense in reducing the challenge to the passive safety-related systems are classified as AP600 Class D. 'Ihis classification conesponds to Regulatory Guide 1.26 Quality Group D. The application of quality standards identified by Regulatory Guide 1.26 and the quality assurance requirements for SSCs that are classified as AP6(X) Class D are descnbed in the response to RAI 260.12.

SSAR Revision: NONE WO5tiflgh0USB

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NRC REQUEST FOR ADDITIONAL INFORMATION Hui Hi!.

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_ i Question 260.14 Generic Ixtier 85-06, " Quality Assurance Guidance For ATWS Equipment That is Not Safety-Related," provides specific quality assurance guidance required by 10 CFR 50.62. Provide a commitment to meet this guidance or describe some other way of rnecting this requirement for the AP600.

Response

The quality assurance guidance provided in Gercric Ixtter R54 will be met for the diverse actuation system, as documented in WCAP 13856, "AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systerns Process."

In addition, the diverse actuation system is classified as an AP600 Class D system. The quality assurance requirements and quality standards for SSCs that are classified Class D are applicable to the diverse actuation systern.

Refer to the response for RAI 260.12.

SSAR Revision: NONE i

! 260.14-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.15 Descnbe the quahty assurance program for fire protection for the AP600 that meets BTP CMEB 9.5-1. i Response; The quality assurance program guidelines of BTP CMEB 9.5-1 are considered and appbed to the design of fire protection systerns protectmg safety-related plant areas and to the preparation of associated comporrnt specificatior-A cornmitment to the application of BTP CMEB 9.5-1 guidelines to procurement, construction and plant operation activities is made in SSAR Subsection 9.5.1.6. Implementation is considered to be the responsibility of the Combined License applicant. BTP CMEB 9.5-1 cnteria C (Control of Purchased Material, Equipment, and Services).

D (Inspection), E (Test and Test Control), F (Inspection. Test, and Operatmg Status), and G (Nonconforming items) are not considered applicable to design cenification.

The extent to which appheable quality assurance requirements are applied to fire protection systems arx1 compotents that protect safety-related plant areas is consistent with the guidelmes of BTP CMEB 9.5-1.

SSAR Revision: NONE l

260.15-1 3 Westinghouse 1

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r NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.16 Justify the exclusion of specific quality assurance program requirements (Ap600 Class D) for items in Table 3.2-1 of the SSAR that are designated as seismic Category 1.

Response

De combination of a nonsafety-related equipment classification and seismic Category I is applicable to a small nuinber of AP600 components. In conformance with the guidance of Regulatory Guide 1.29, seismic Category I components, including the limited number that do not have a safety-related function, meet the quality assurance requirements of 10 CFR Part 50 Appendix B. The combination of a nonsafety-related equipment classification and seismic Category I for a limited number of items is consistent with the practice in operating nuclear power plants.

He items with a nonsafety-related equipment classification and seismic Category I classification are equipment related to fuel storage, integrated head package, and reactor internals that are not core supports. In each of these areas designation of a nonsafety-related classification does not exclude quality asurance program requirements.

The design specification and functional specification requirements for these items specify the appropriate quality assurance requirements.

SSAR Revision:

Revise the fourth paragraph of Subsection 3.2.1.1.1 as follows:

Seismic Category 1 Safet :'-: d :tructures, systems, and components meet the quality assurance require-ments of 10 CFR 50, Appendix B. He criteria used for the design of seismic Category I structures, systems, and components are discussed in Section 3.7.

260.16-1 W -

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r NRC REQUEST FOR ADDITIONAL INFORMATION i -

Question 260.17 State whether the quahty assurance requirernents for construction are considered a COL responsibility by Westmghouse.

Response

The quality assurance requirernents for constmetion are the responsibihty of the Cornbined License applicant.

SSAR Revision: NONE W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 260.18 Desenbe the quality ass- ts ttut are being applied to software for the design phase of the AP600-

Response

Quabty assurance provisions for software used m safety-related applications are descnbed in WCAP-8370 Revision 12 A Part B Section 5. These provisions address software development, software control and documentation, software testing, and software procurement.

Quality assurance provisions for safety-related software meet the requirements of ASME NQA-2a-1990 Part 2.7, and are documented in written procedures that address the software development process, verification and validation of computer programs, configuratwn control of computer programs, software enor reporting and resolution, dedication and installation of external computer programs, and single application and small intemal-use computer programs.

WCAP 13383 Revision 0, "AP600 Instrumentation and Control Hardware and Software Design, Verification, arx1 Validation Process Report," further desenbes the implementation of the software quality assurance program for I&C software.

Software quahty assurance audits are performed to evaluate the effectiveness of these prc 'sions and to identify corxiitions requinng corrective action.

SSAR Revision: None l

260.18-1 W=

W85tingl10US0 i l

NRC REQUEST FOR ADDITIONAL INFORMATION

- ,1-Question 260.19 Desenbe what quality assurance provisions apply to the design phase test programs for the AP600.

Response

Quality assurance provisions for test pmgrams used for safety-related appbcations are described in WCAP-8370 Revision 12A Part B Section 11 and address test planning arxl prerequisites, test requirements, test procedures, and test results and records.

The organizations that execute the AP600 test programs are considered to be suppliers of testing services. The selection and evaluation of suppliers, control of changes in services, acceptance of roervices, and control of supplier nonconformances are performed in accordance with the requirements of NQA-1 Cnterion 7. For those cases in which testing is performed by AP600 Test Engineenng personnel, a Project Quality Plan has been written to implement appropriate quality requirements.

Audits are performed to evaluate the effectiveness and implementation of the QA program provisions for testing activities, arxl to identify conditions requiring corrective action.

SSAR Revision: None 260.19-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 ilu- H 3 p..

A Question 410.23 Section 5.4.11.2 of the SSAR states that the discharge of water, steam, and gases from the first. stage automatic depressurization system valves when used to vent noncondensable gases does not result in pressure in excess of the in-containtnent refueling water storage tank (IRWST) design pressure.

Wben the high pressure discharge of steam and noncondensable gases, through piping system, injects into the in-contaimnent water storage tank, it will be condensed and mixed with the tank water. The IRWST, in this case, will function similarly to the suppression pool in tbc BWR plants. Provide an analysis to demonstrate that the hydrodynamic loads on the tank and piping have been adequately considered.

Response

The hydrodynamic loads imposed on the IRWST by manual venting of non-condensable gases using the first stage automatic depressurization system (ADS) valves are bounded by tbc analysis perfonned during operation of the automatic depressurization system for sequential operation of stages one to three which vent steam and water to the in-containment refueling water storage tank (IRWST).

The analysis method for the hydrodynamic loads on the IRWST is described in SSAR Subsection 3.8.3.3.2.

Dynamic loads on tbc IRWST due to automatic depressurization system operation are determined using tbc results from the automatic depressurization system hydraulic test described in WCAP 13342. This test determines tbc dynamic effects on the test tank by simulating operation of tbc automatic depressurization system and spargers.

Loads on the IRWST boundary are calculated using the pressure source load obtained from the automatic depressurization system test. Hydrodynamic finite element analyses are performed on both the test tank and the IRWST. The analyses on the test tank develop the source pressure load by comparison of pressure measurements at selected locations in the tank against results at tlase locations predicted by the analyses. Tbc source pressure load is then used in the IRW71' analysis to give the dynamic ;esponses of the tank boundary (deflections, accelerations, and stresses). Appendix 3F will be added to the SSAR describing these analyses. This appendix is proprietary and is provided under separate cover.

The significant hydrodynamic loads on the ADS discharge piping are the thrust reaction loads due to Guid exiting through the sparger holes and tbc internal unbalanced pressure and momentum loads on each straight pipe segment.

The pressurization of the sparger arm is analyzed using a detailed W-COBRATRAC model. The results of this analysis are used to calculate the thrust loads at the sparger holes.Tbc internal unbalance pressure loads are obtained from the ITCH computer program"). This program has been benchmarked to EPRI test data for pressurizer relief valve discharge loadings. These hydrodynamic loads are calculated based on sequential operation of ADS stages one to three which vent steam and water. *Itese loads envelop tbc manual venting of non-condensable gases using the first stage automatic depressurization system (ADS) valves.

Reference:

1) WCAI .9924, ' ITCH valve code description and verification *, M. A. Berger and K. S. Howe, July 1982 410.23R1-1 3 Westinghouse Rev sion 1

NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 1 M ,

SSAR Revision:

Add Appendix 3F as provided under separate cover.

Revise paragraph 2 of 3.8.3.3.1 as follows:

Dynamic loads on the IRWST due to ADS operation are determined using the results from the ADS hydraulic test-dese+ibed h Sr.t ; L&. This test simulates operation of the ADS and spargers, and determines the dynamic effects on the test tank. Loads on the IRWST boundary are calculated using the pressure source load obtained from the ADS test. Hydrodynamic finite element analyses are performed on both the test tank and the IRWST. The analyses on the test tank develop the source pressure load by comparison of pressure measurements at selected locations in the tank against results at these locations predicted by the analyses. The source pressure load is then used in the IRWST analysis to give the dynamic responses of the tank boundary (dellections, accelerations, and stresses ). Additional information on these analyses is provided in Appendix 3F.

410.23 R1-2 W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION ali Response Revision 1 Ouestion 440.13 In Section 5. A, item D (p. 8) of Revision 0 to WCAP-13342, a vacuum breaker is indicated as being provided on the discharge piping, above the quench tank (simulated IRWST) water level, to prevent backDow into the ADS discharge line from the IRWST. Is this vacuum breaker prototypic with respect to the AP600, or is it provided only for the ADS test? If it will also be included in the plant, its design and operating conditions should be established and tested adequately. No specific information on the vacuum breaker design nor on how it wiil be tested is included in WCAP-13342, nor is any such component indicated in the description of the ADS in Chapter 6 of the SSAR. Provide detailed information of the vacuum breaker design.

Response

The AP600 ADS discharge piping that connects from the pressurizer to the spargers includes a vacuum breaker downstream of the associated ADS valve group. Dese vacuum breakers are shown on the piping and instrumentation drawing in the SSAR, Proprietary Volume 2, page P5.1-9. Subsection 5.4.6 and Section 6.3 of the SSAR provide information about the automatic depressurization system arrangement and operation, but do not specifically discuss the vacuum breakers. Additionalinformation about the vacuum breakers as shown below will be added to SSAR Subsection 5.4.6.2.

Vacuum breakers are provided in the AP600 ADS discharge lines to help prevent destmetive water hammer following ADS operation. De vacuum breakers limit the pressure reduction that could be caused by steam condensation in the discharge line and thus limit the potential for liquid backflow from the in-containment refueling water storage tank following ADS operation.

He vacuum breaker installed in the ADS test facility is not a prototype for the vacuum breaker that will be installed in the ADS discharge lines for the AP600. The vacuum breaker included in the design of the test facility prevents excessive water from being drawn back into the sparger discharge line following the various ADS tests to protect the discharge line.

He perfornumee of the vacuum breaker in the ADS test facility will be evaluated to gain an understanding of the vacuum breaker operation during the exter sive range of test conditions. De results from the ADS testing will then be used to help develop the detailed vacuum breaker design requirements and specifications for the actual vacuum breakers for the AP600. The results of the vacuum breaker performance during the ADS testing, along with the analyses and evaluations used to develop the design specifications for the plant vacuum breakers, will be sufficient to conservatively bound the expected design and operating conditions without additional testing.

SSAR Section 5.4.6.2 will be revised as follows:

WC5tingt10t!SS

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 SSAR Revision:

For each discharge path a pair of valves are placed in series to minimize the potential for an inadvertent discharge of the automatic depressurization system valves. Tlie fourth stage valves aad operators are designed such that these valves cannot open against full reactor coolant system normal operating pressure.

Vacuum breakers are provided in the AP600 ADS discharge lines to help prevent water hammer following ADS operation. The vacuum breakers limit the pressure reduction that could _be caused by steam condensation in the discharge line and thus limit the potential for liquid backflow from the in-containment refueling tvater storage tank following ADS operation.

440.13R1 -2 W

Westingflouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.36 The Reactor Coolant System tRCS) Piping and Instrumentation Diagram (P&lD) on p. P5.1-9 of Chapter 5 of the SSAR shows, on the parallel Automauc Depressurization System ( ADS) valve groups (stages 1-3). additional solenoid. operated valves, V007A. -B, -C, and -D. The purposes of these salves do not appear to be described m the SSAR.

a. What are the functions of these valves 7

b. It appears fmm the P&ID that the normal state of these valves is " closed." Under what conditions do the valves open. and is the operation automatic or manual?

c. What happens to the valves on the generation of an "S" signal and on actuation of the ADS 7

d. Has this configuration been analyzed to determine whether it is possible to create an unisolated leak from the pressurizer, beyorx1 the capability of normal makeup? If so, present the analysis of the eveut; if not, perform the analysis or justify the asswnpuon that an unisolable leak cannot occur.

c. In the Failure Modes and Effects Analysis, Table 6.3-6 of the SSAR, p. P6.3-51, and on the P&lD for the Passive Core Cooling System, Figure 6.31 of the SSAR, p. P6.3-59, valves V007A/B are identified as "pressunzer to CMT line check valves." His definition appears to be inconsistent with the RCS P&lD cited above. Reconcile this inconsistency.

f. In the Failure Modes and Effects Analysis, Table 6.3-6 of he SSAR, p. P6.3-51, valves V002A/B and V003A/B are identified as "CMT inlet and outlet isolation AOVs," while on p. P6.3-57 of the same table. these valves are identified as ADS MOVs, The same inconsistency is carried through on the associated P&lDs for the Passive Core Cooling System and the Reactor Coolant System.

Reconcile this inconsistency.

Response

a. Rese valves are provided to facilitate IST of the ADS valves. They will be used in two different modes.

One mode is to support zero differential pressure periodic IST during power operation. The other mode is to support partial differential pressure IST during refueling operation.

For periodic IST, these test valves are opered to ensure that there is no differential pressure across the ADS valves when they are stroke tested. For example,in order to test ADS valve V002C, test valve V007C is first opened. His equalizes the pressure downstream of V002C and the RCS. This allows stoke testing of the ADS valves with reduced wear on the valve internals and contspondingly reduced risk of leakage.

440.36-1 1 3 Westingh0uSe j l

NRC REQUEST FOR ADDITIONAL INFORMATIC' sE

~.

For refueling shutdown IST. these test valves are opened to ensure that a pre-detennined differential pressure is applied across the ADS valves when they are tested. This provides a partial differential pressure

/ low How test. For example, in order to test ADS valve V002C, test valve V007D is first opened. This equalizes the pressure downstream of valve V002C and the IRWST, resulung in a differential pressure across V002C equal to the RCS pressure. The RCS pressure to be used in this test is expected to be in the range of 400 to 1000 psig. As the ADS valve opens the differential pressure will transfer to the test line because it is much more restrictive.

b. These valves are normally closed. The only time they are anticipated to be opened is during ADS valve IST as discussed in item a. These valves do not have any autornatic controls.

c. As discussed in item b, these valves do not have any automatic controls.

d. Unisolable leaks through tirse test valves will be less than the CVS makeup capability because these test lines contain 3/8" flow restnctors. These restrictors are required by notes on the RCS P&ID, SSAR figure 5.1-5.

e. These valve numbers are not inconsistent. Each system starts numbering valves at V001. The full valve identification number also contains the system abbreviation and a component type code (PL) which is not shown on the P&lDs but is identified on each P&ID in note 1. For example, the full identification number for ADS valve V002A is RCS PL V002A and for CMT inlet isolation valve V002A is PXS PL V002A.

f. These valve numbers are not inconsistent as drscussed in item c.

SSAR revision: NONE i

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l 440.36-2 !

W-Westinghouse !

NRC REQUEST FOR ADDITIONAL INFORMATION gli - fiig Question 440 37 Table 63 5 of the SSAR, p. P63-49. lists pararneters for the passive RHR heat exchanger. The product of the overall heat transfer coefficient and heat transfer area (UA) is pven as 3.80 x 10' BTU /hr F.

a. On what heat transfer area (tube inner surface on tube outer surface)is the UA product based, and what is that area?

b. Are the numbers for UA and for the design heat transfer for each heat exchanger or for both heat exchangers taken together?

c. What are the individual heat transfer coefficients on the inside and outside of the heat transfer tubes, and how were they calculated?

d. What are the tube tiermal conductivity, atxi inner and outer diameters?

e. What provisions in the UA product were made to account for scale on the tube inner and outer surfaces?

f. Justify the use of 120*F as both the inlet and outlet temperature on the shell side of the heat exchanger.

g. The outlet temperature on the primary (tube) side of the heat exchanger is shown as 224 F.

However, it is stated in Section 63.2.2.5 of the SSAR, p. P63-9, that the vertical run of the C-tubes is approximately 18 ft. At a submergence of about 18 ft., assuming the containment is at atmospteric pressure, the static pressure at the bottom of the tube would be approximately 21.9 psia. From steam tables, the saturation temperature at 21.9 psia is approximately 230 F. Is tir shell side coolant in surface (subcooled) boiling along the entire length of the C-tube? If so, explain how the tube side temperature at that point can be less than the saturation temperature of the stell side coolant. Provide a table or plot of shell-side and tube-side water temperatures as a function of length along the C-tubes.

Response

The value provided in Table 63-5 is an approximate UA for the PRHR heat exchanger, based on its expected performance predicted by the analysis codes during an accident. The analysis codes do not use the overall UA presented, but ratirr use detailed beat transfer correlations. See the response to RAI 44035 for a detailed explanation of the heat transfer correlations used to predict the performance of tte PRHR heat exchanger. The UA is presented only to provide an estimate ofits heat transfer performance, at the given conditions.

a. The UA is not based on tube surface area. See the paragraph above.

[ We5tingfl0USB

NRC REQUEST FOR ADDITIONAL. INFORMATION

_ ;;g

!E

b. The UA and design heat transfer applies to caly one heat exchanger. As descnbed in Chapter 15, only one PRHR heat exchanger is assumed to be available for heatup events. and other events when PRHR operation is beneficial. For osercochng transients (such as steam line breaks) both PRHR heat exchangers are assurned to operate.

c. See the response to RAI 440.35 for an explanation of the heat transfer correlauons used to predict the performance of the PRHR heat exchanger.

d. 'The following are design parameters for the PRIIR tubes:

Matenal SS-316L ID / OD (inches) 0.62 / 0.75 Thermal Conductivity See the response to RAI 440.35

e. The heat transfer correlations do not specifically account for scale on the inner or outer tube surface area. Conservatism has been incorporated into the analysis codes, considering the conservative minimum heat transfer correlation developed from the test, as well as tie assumptions regarding the number of tubes plugged and the number of heat exchangers avadable.

f. 'Ihe 12(TF refers to the IRWST temperature at the time when the otter parameters (inlet temperature, flow rate) are expected to occur.

g. The tube side outlet temperature presented in this table corresponds to an IRWST temperature of 120"F. 'the PRHR heat transfer is dominated by two mechanisms. When the tube wall outside temperature is n saturation nucleate boiling occurs on the outside tube surface. As the primary side temperature is reduced along the length of the tubes, and the outside wall surface temperature is reduced to below saturation, heat transfer is dominated by free convection on tie outside of the tubes. Wien the IRWST is subcooled, it is possible for the primary side outlet temperature to be less than the IRWST water saturation temperature. The heat transfer mechanism under tiese corxlitions is free convection.

SSAR Revision: NONE l

440.37-2 W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION awe c

Question 440.38 Clanfy the operation of the hydrogen addiuon system and the provisions for venting hydrogen that accumulates in the pressunzer during normal pl ant operation. In addition, explam the impact of these operations on the performance of reactor and containment systems, and on plant accident response. Specifically:

a. The text of Secuon 9.3.6.1.2 of the SSAR and the associated figmes of the Chernical and Volume Control Sys!cm indicate that hydrogen for oxygen control is introduced directly into the RCS, maintaining a concen: ration of 25-35 cc per kg (at STP) using a gas cyiinder. Confirm that this is ite case.

b. Specify the projected losses and consumption of hydrogen that comprise the specified rate of addition of 4000 cc per minute.

c. Section 5.6.5 of the SSAR describes operation of the pressurizer and associated subsystems. The spray subsystem operates constantly to balance boron concentration and to keep the spray piping hot. The coolant that is sprayed into the pressurizer will be stripped of much of its dissolved hydrogen, with this gas building up in the pressurizer. Degassing of the pressurizer is descnbed as being required only on an occasional basis, and is accomplished using the stage i valves of tie automatic depressunzation system, How often is it projected that degassing of the pressurizer wdl be required, and what is the basis for selecting that frequency of operation?

d. What is tie volume of hydrogen projected to be removed from the pressurizer during each degassing operation?

e. The hydrogen vented from the pressurizer will flow through the ADS sparger into the in-containment refueling water storage tank (IRWST). What actions will be taken to avoid a buildup of hydrogen above the IRWST or in the contauunent that could eventually result in a combustible concentration of the gas?

f. What is the in-containment radioactive source that results from the venting of hydrogen from the pressurizer?

g. As hydrogen builds up in the pressurizer, it will also accumulate in tir piping associated with the passive safety injection systems, such as pressure balance lines. How will accumulation of hydrogen affect the performance of the passive SI systems? Present analyses to support your evaluation.

h. How has tic use of the ADS Ist stage valves for pressurizer degassing been accounted for in assessing the likelihood of inadvertent ADS actuation as an accident initiator?

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

Response

a. The AP(MU mamtains hydrogen wittun design limits by injecting hydrogen directly into the RCS via the CVS retum line. These operations will be performed on a batch mode basis as required.

The source of hydrogen is gas cylinders connected to the CVS piping.

b. The design basis consumption rate of hydrogen in the RCS (specified as 4000 cc/ min) is a conservative estimate based on previous Westinghouse experience. This estimate includes hydrogen consumed by combining with fme oxygen, leakage through the steam generator tubes, valve packing, etc., and expected RCS makeup operations to account toi ECS leakage and nomul dilutions. It is expected that the actual consumption rates will be less than that of cur ent plants due to fewer expected leakage paths, and infrequent letdown operations.

c. Section 5.4 refers to the first stage ADS valves being used to vent non-condensible gasses from the pressurizer. 'Ihis operation will only be required following an accident. Revision I of the AP600 SSAR clanfies this operation.

It is not expected that degassing of the pressurizer will be required during normal operations.

Westinghouse experience has shown that routine pressurizer venting is not necessary, even in plants that frequently operate spray. As described in Section 5.4.5 of the SSAR, a small continuous spray flow (referred to as spray bypass flow)is provided to prevent excessive cooling of the main spray piping, and to minimize differences in boron concentration in the pressunzer. This spray bypass flow is very small (2-4 gpm) and is similar to the spray bypass flow rates of current Westinghouse plants. However, based on plant experience, even plants that frequently run normal spray have not required ventmg of hydrogen from the pressurizer.

Ahhough venting of hydrogen from the pressurizer is not anticipated, the Primary Sampling System could be used to perform this function. A sampling connection is provided in the pressurizer steam space that could be aligned to vent to the Gaseous Waste Processing System (WGS). The first stage ADS valves would not be used to perform this function.

d. It is not projected that bydrogen will be required to be vented from the pressurizer.

e. The first stage ADS valves will not be used to vent non-condensibles frorn the pressurizer during normal operation.

f. This is not a concern since the first stage ADS valves are not be used to vent non-condensibles from the pressurizer to the IRWST during normal operation.

g. The effects of non-condensible gasses are included in the Waltz Mill CMT test. These test results and analyses wdl massive the effects of non.condensibles on CMT performance.

1 i

440.38-2 3 Westinghouse 1

1

NRC REQUEST FOR ADDITIONAL INFORMATION

h. The first stage ADS valves will not be used to vent non-condensibles frorn the pressurizer during normal operation.

SSAR Revision: NONE 4

38-3 Vj Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

'iii~ iii F ?

e Question 440.39 Clarify the meaning of "fadun: of a CMT discharge valve" discussed in tir analysis of a steam system piping fadure in Section 15.1.5.2.1 of the SSAR. Does this mean that the CMT tlow is reduced or does it mean that a single CMT is assumed to te unavadable?

Response

The statement made in Section 15.1.5.2.1 of the SSAR that: " ..the core makeup tank injection lire characteristics rmxieled reflect the failure of ore core makeup tank discharge valve.. " is interkled to indicate that the CMT flow is reduced. The injection line from each CMT contains two air operated, fad open valves connected in parallel. The maxirnum amount of CMT injecuan flow occurs if both these valves are open Should one of these valves fail to open, the parallel vdve provides a path for the injection flow. With only the single valve open the injection flow is reduced.

SSAR Revision: None W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION sp i!!E Ouestion 440.40 The arulysis of a steam system piping failure presented in Section 15.1.5.2.3 of the SSAR discusses 10 secorxis of flow from the unaffected steam generator before the mam steam isolation valves close. Is this 10 seconds of flow meluded in the transient analysis?

Response

The ten seconds of main steam flow from the unaffected steam generator is mcluded in the transient analysis. The analy sis explicitly models the steam flow out the break from the " intact" steam generator, that occurs prior to closure of the mam steam isolation valves.

SSAR Revision: None W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION iiiE EM Question 440 41 In the analysis of a steam system pipmg failure presented in Section 15.1.5.2.4 of the SSAR, the departure from nucleate bothng ratio (DNBR)is discussed, but no figure is provided to show results for the hmiting case. Provide such a figure.

Response

Westinghouse has not presented DNBR versus time plots for the steam system pipmg failure event (main steamline break)in the Final Safety Analysis Repon for any plant. The AP600 SSAR results reponed simply renect tius fact.

The main steamline break is a Condition IV event and tirrefore must meet the radiological dose release requirements of 10CFR100. Limited fuel damage is permitted as long as the above criterion is met. However, Westingbouse has conservatively applied the Condition Il cnterion that tir DNB design basis be met for the main steamhne break.

The general methodology employed to analyze this accident involves the use of the LOFTRAN computer code to predict limiting plant charateristics as a function of time. Specifically, the parameters involved are RCS pressure, cose inlet temperatures, core flow rate, and core boron concentration. Detailed nuclear design codes then use the LOFFRAN results to predict the core reactivity. The combination of the LOFTRAN results with tie nuclear design code predictions then produces a limiting "statepoint" that is subjected to detailed thermal-hydraulic analysis. Several different sets of conditions are examined in the vicinity of this limiting statepoint, but no continuous set of DNBR calculations is performed over the entire ler.gth of the event.

Using the statepoint methodology, minimum DNBR calculations have been performed for the AP600 main steamline break; transient. For a conservative bounding case, the predicted minimum DNBR has been determined to be greater than 3.0. These results demonstrate that the DNB design basis is met. No fuel failures are predicted for this transient.

SSAR Revision: None W W85tinEh0Use

NRC REQUEST FOR ADDITIONAL INFORMATION E

4..pp Question 440.42 in the analysis of a feedwater system ppe break presented in Section 15.2.8.2.1 of the SSAR, a 17 second delay is assumed for passive residual heat removal (PRHR) system actuation. What is the rationale for this assumption?

Is the assumption conservative?

Response

The 17-second delay referenced in Secuon 15.2.8.2.1 of the SS AR is not correct. In the feedwater system pipe break analysis presented in the SSAR Section 15.2.8, a time delay of 20 seconds was used for actuation of the passive residual heat removal system. The corrected SSAR text is shown below.

ne PRHR contains valves in the outlet piping which are nonnally closed. Wien a signal to actuate the PRHR occurs these valves open to allow How through the PRHR. The tune delay to fully open these valves is the time delay mentioned in 15.2.8.2.1. The best estimate time to fully open the PRHR valves is 10 seconds. For feedwater system pipe break events which result in a heat up of the RCS it is conservative to delay the actuation of the passive residual heat removal system. Therefore, a conservatively long delay of 20 seconds was used la the feedwater system pipe break analysis.

SSAR Revision:

15.2.8.2.1 Method of Analysis

. The passive residual feat removal system is actuated by the low steam generator water level (wide range) signal. A 20-secorx1 delay is assumed following the low level signal to allow time for tie alignment of tie passive residual heat removal valves.

W WestinEhouse ,

I

NRC REQUEST FOR ADDITIONAL INFORMATION jrE El{i Ouestion 440.43 In the analysis of the locked rotor event presented in Section 15.3.3.3 of the SSAR, an assumption is made for the analysis of radiological consequences that 18% of the fuel rods fail. According to the Standard Review Plan (SRP) guidance for this event, fuel rod fadure is to be assumed for any rod for which DNB is predicted to occur. Provide results of the analysis of this event that demonstrate that the assumed fuel failure percentage is consistent with SRP requirements.

Response

For the locked rotor event presented in Section 15.3.3 of the SSAR,18% of the fuel rods were calculated to undergo DND. Therefore, the assumed fuel failure percentage is consistent with the Standard Review Plan.

SSAR Revision: None W85tingh0US8 l

1

NRC REQUEST FOR ADDITIONAL INFORMATION

=

l Question 440.44 Provide a rationale for the assumpnon that the locked rotor event is most limiting for the section addressing decrease in reactor coolant system flow rate. (Chapter 15 3)

Response

Section 15.3 of the SSAR presents analysis for transients which result in a decrease in reactor coolant flow.

Analyses are performed for the following specific events; Partial Loss of Forced RCS Row (Section 15.3.1)

+ Complete Loss of Forced RCS Flow (Section 15.3.2)

RCP Shaft Seizure (Locked Rotor)(Section 15.3.3)

RCP Shaft Break (Section 15.3.4)

The RCP shaft seizure analysis presented is postulated as an instantaneous seme of the RCP rotor. This results in a rapid decrease in RCS flow (see SSAR Figures 15.3.3.l A & 15.3.3-1B). The itCP shaft break is postulated as an instantaneous break of the shaft between the pump rotor and the motor. This results in a rapid decrease in the flow similar to the locked rotor event, however the rate of the flow reduction is not as great as that of the shaft seizure event. The partial and complete loss of RCS flow events are postulated to occur due to RCP motor or power supply faults. Dunng these events the inertia of the faulted reactor coolant pumps and motors retards the reduction in RCP ; peed and RCS flow decreases more slowly (see SSAR Figures 15.3.1-I A and 15.3.2 1) than the locked rotor event .

All of the above transients are promptly mitigated by a reactor trip on low reactor coolant flow or low reactor coolant pump speed. The rate of flow reduction prior to reactor trip is greatest during a locked rotor which results in tir most severe power to flow nusmatch transient. Therefore a locked rotor results in a more severe reactor coolant flow transient chan other events which result in a reduction in RCS tiow.

SSAR Revision: None W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION g _

2 Question 440.45 In the analysis of a rod ejection event presented in Section 15.4.8.3 of the SSAR,159 of the fuel rods are assumed to fail. This is contrasted to the analysis in Section 15.3.3.3, w here 189 of the f uel rods are assumed to f ail. What is the rationale for the difference in these assumptions?

Response

The basis for the number of failed fuel rods for both the RCCA Ejection event presented in Section 15.4.8 of the SS AR and the Locked Rotor event presented in Section 15.3.3 of the SSAR is the number of fuel rods in DNB.

As stated in the response to RAI 440.43. the 18M fuel failures for the Locked Rotor accident was detennined as a result of a specific calculation performed for that event. The basis for the number of fuel failures in the RCCA Ejection event is presented in Section 15.4.8.2.1.8 of the SS AR,in which the 15% was conservatively chosen based on the referenced analysis for that event. It is not expected that the two transients would result in the same number of fuel rods in DNB or consequently the same number of rods assumed to fail for the detennination of the offsite dose consequences.

SSAR Revision: None 1

l 440.45-1 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

=

Ouestion 440.46 in the analysis of the inadvertent operation of the ADS presented in Section 15.6.1.1 of the SSAR, it is stated that "the design of the fourth ADS stage valves is such that the valves cannot be opened while the reactor coolant system is at nominal operating pressure." However, Westinghouse has stated in meetings with the NRC staff that a fmal design for the fourth stage valves has not yet been determined, and no test data currently exist for these valves demonstrating the stated characteristics. Accordingly, present an analysis for the inadvertent opening of one fourth stage ADS stage, or show that the consequences of this event are bounded by other hot leg LOCA analyses.

Response

The AP600 SSAR presents a double-ended hot leg guillotine break analysis. For this break the peak cladding temperature (PCT) occurs at time zero, during full power steady-state operation. The shutdown of the core due to voiding together with the exceptional core cooling obtained as flow proceeds in the normal direction through the core to the hot leg break location prevents any cladding heatup; the cladding temperature barely exceeds the core fluid saturation temperature during the entire transient. The break area for this case (5.24 sq. ft. is the cross-sectional area of each side of the break) is far larger than the vent area fless than one sq. ft.) associated with the inadvertent opening of one ADS fourth stage flow path.

Any smaller postulated hot leg break will also exhibit its calculated PCT value at time zero (steady-state operation).

The inadvertent ADS actuation transient presented in the SSAR is representative of smaller hot leg break cases.

Core shutdown is achieved via control rod insertion, and the action of the AP600 passive systems maintains the reactor vessel inventory well above the top of the active fuel. he PCT once again occurs at time zero of this postulated break case.

The non-limiting nature of hot leg breaks not only holds for the AP600 but also is typical of standard Westinghouse PWR designs. Hot leg break locations have been generically shown to be non-limiting in both large and small break ECCS Evaluation Model computations. Consequently, hot leg breaks are not normally analyzed as part ofindividual plant licensing submittals. There is no need to consider postulated hot leg break events beyond those described above which already exist in the AP600 SSAR. His holds true independent of the specific fourth stage ADS valve characteristics.

SSAR Revision: None PRA Revision: None 440.46-1 W.-

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION g~ .=3 e

Question 440.47 In the analysis of a steam generator tube rupture presented in Section 15.6.3.2 of the SS AR, steam generator overfill is discussed as pan of the event sequence. The assumpnons appear to be reasonable, but no analytical results are shown (e g., what calculatmnal method). Provide these results.

Response

Consistent with the steam generator tube rupture (SGTR) sections in the FSARs of Westinghouse operating plants.

and the NRC Reg. Guide 1.7 Rev. 3 guidelines for format and content, the SGTR analysis presented is the design basis SGTR which is limiting with respect to the radiological dose releases. Sensitivity studies performed to determine the hmiting case for dose releases (conservative single failure etc.) as well as the analyses used to determine that the ruptured steam generator will not overfill are referred to in the SAR but are not presented m detail.

As noted in Section 15.6.3.2 of the SSAR an analysis was performed to demonstrate margin to steam gene.ator overfill in the event of a steam generator tube rupture, without taking credit for expected operator actions. The analysis assumed reactor trip, chemical and volume control system actuation and core makeup tank initiation at the start of tie transient. Reactor tnp is assumed at the start of the event because feedwater to the ruptured steam geterator is assumed to be throttled prior to reactor trip, so that the combined flow into the steam generator matcles steam flow out. Thus, the steam generator level will not increase until reactor trip. With a reactor trip at the start of the event the primary pressure remains high, resulting in a high rate of break flow at the time when the steam generator level will begin to increase. The CVS pumps actuate and CMT injection is initiated early to attempt to maintain the pressure at a high value and increase break flow. The cold CMT injection removes some of the decay heat resulting in reduced steam generator PORV steaming and less mass removal from the ruptured steam generator.

The single failure (failure of the stanup feedwater control valve to throttle flow when the nominal steam generator level is reached) and other conservative assumptions employed in the analysis (such as high initial steam generator secondary level, maximum initial RCS pressure, offsite power available, maximum CVS injection flow, maximum stanup feedwater flow and minimum stanup feedwater delay time) are noted in Section 15.6.3.2, The analysis w.u perfonned with the LOFITR2 program described in Section 15.6.3.2.1.1 of the SSAR. The sequence of events for the overfill analysis is presented with selected plots on the following pages. The analysis demonstrates that ttere 2

is 593 ft' margin to the maximum steam generator volume of 5554 ft .

SSAR Revision: None i

440.47-1 W-Westinghouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION EVENT Time (seconds)

Double ended steam generator (SG) tube rupture O Reactor tnp and main feedwater pumps begin coastdown 0 Chemical and volume control system (CVS) pumps actuated 0 Startup feedwater flow initiated 20 Core makeup tank (CMT) injection initiated (following delay) 22 CVS and stanup feedwater pumps isolated on high-2 SG narrow range setpoint and 784 passive residual heat removal system (PRHR) initiated on high-2 SG narrow range level setpoint Maximum ruptured SG water level (4%1 ft') reached and break Dow terminated 2364 440.47-2 W Westinghouse

. _ . . , .. _ . . . . _ _ _ .- . m... ._ .. . _ . 2. 3 f

i NRC REQUEST FOR ADDITIONAL INFORMATION i

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i so ..!

40 l u ;

h 30 I s ,

, -r 3 20.

f 8 (

to .

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i o, 1000 2000. 3 00. 4000. 5000. 6000. 7000 l Iset (MC) ;

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. Margin to Steam Generator Overfill: Break Flow and Ruptured Steam Generator Water Volume

(

a l

440.47-3 :

W westinghouse .i 1

i l

NRC REQUEST FOR ADDITIONAL INFORMATION e:m ::

T "@

Ouestion 440.48 The analysis of a single steam generator tube rupture (SGTR) event is presented in Section 15.6.3 of the SSAR, with a detailed desenption of the assumptions and results in Sections l$.6.3.2.1.2 and 15.6.3.2.1.3. Among the assumptions listed in Section 15.6.3.2.1.2 is the availability of offsite power or onsite ac power, since both the chemical and volume control system (CVS) and startup feedwater system (SFWS) are assumed to be available.

These assumptions are inconsistent with the guidance in the Standard Review Plan to include corsideration of a loss of offsite power. They are also in conflict with Commission guidance on the use of non-safety equipment for hmiting design faults, and with the EPRI Passive Plant Utility Requirements Document (URD), Volume III, Chapter 5, Paragraph 1.2.2, that states that only safety-related equipment is assumed to be available for LDB events.

In addition, Westinghouse's analysis includes opening the steam generator PORV, which appears to te in conflict with Paragraph 4.2.5 of Chapter 3 of Volume III of the URD, which indicates tuat passive plants should be able to sustain a single SGTR without lifting steam generator relief valves. Provide the following infomtation:

a. The staff understands that the assumption of CVS flow tends to increase the primary-to-secondary break flow, thus maximizing inventory loss from the primary system. However, makeup from the CVS also delays the injection of makeup flow from the core makeup tanks (CMTs), and thus reduces the possibility that the CMTs will dmin to the automatic depressurization system (ADS) first stage actuation setpoint. Present a new analysis assuming no onsite or offsite ac power availability to demonstrate that the AP600 is capable of being brought to a stable condition with (a) no operator action and (b) use of passive safety systems only.

b. Why does the AP600 design comply with the EPRI URD requirement for no steam generator relief valve actuation in the event of a single SGTR?

c. The staff notes that Westinghouse's SGTR analysis requires flow from the secondary system back into the primary to stabilize the system near the end of the event. Describe how the effects of boron dilution on reactivity that can result from this backflow are accounted for in the anslysis medel.

Response

a. The requested analysis is currently being performed. Results will be provided by March 15, 1994.

b. The EPRI URD requires that the plant be capable of accepting a steam generator tube rupture without lifting steam safety valves. His requirement is met since the combination of plant trip and turbine bypass actuation will prevent the steam side safety valves from lifting. The URD states that for stears generator tube rupture events without assumed operator actions water relief via the safety valves or PORVs must be prevented on a best estimate basis,i.e. without the turbine bypass system. This har , een demonstrated by the steam generator overfill analysis discussed in Section 15.6.3.2 of the SSAR.

c. The SSAR SGTR analysis was performed to conservatively calculate the offsite radiological doses using the LOFTTR2 program employed in SGTR analyses for currently operating W plants. This analysis model W

- Westin8 house

NRC REQUEST FOR ADDITIONAL INFORMATION EEE 6%

does not track boron with the break flow. A calculation is underway to determine the effects of backflow from the secondary system on reactivity. Results will be provided by March 15, 1994.

SSAR Revision: NONT l

~

W Westinghouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Question 470.9 Provide the doses from a postulated LOCA based upon the release rate and timing of NUREG-1465 t Section 15.6.5).

Response

A new dose analysis has been perfonned based on the curwnt draft NUREG-1465 (June 1992) with the followinc exceptions:

1. As advised by NRC staff at a meeting with EPRI and Westinghouse on May 18,1993, the core releases for the design basis LOCA are tenninated at the end of tie early in-vessel release phase identified in draft NUREG-1465. This assumes that core cooling is regained prior to vessel failure.

2. As advised by NRC staff at a meeting with EPRI and Westinghouse on May 18,1993, the iodine species split will be modified to include organic iodine with 5 percent of the elemental iodine assumed to convert to the organic fonn. Thus,instead of the split of 95 percent as particulate and 5 percent as I and H1 currently speci6ed in draft NUREG-1465, the species split used in the analysis is 95 percent as particulate,4.75 percent as I and Hl. and 0.25 percent as organic iodine.

3. Draft NUREG-1465 specifies that the gap release phase is initiated at 25 seconds into the large break LOCA and that the duration of the gap release is 30 minutes. The AP600 core response is calculated to involve less than 5 percent rod bursts early in the accident. The remaining fuel rods are not projected to burst until more than an hour into the accident. De dose analysis reflects this.

4. The discussion of particulate rernoval coefficients provided in draft NUREG-1465 suggests extremely low rates of removal that am inconsistern with the removal rates calculated for AP600. The dose analysis utilizes a conservative removal coefficient specific to the AP600.

The doses calculated using the draft NUREG-1465 source term (revised as discussed above) result in the following doses:

Site Boundary: Thyroid 88 rem Whole body (immersion) 0.5 rem LPZ Boundary- Thyroid 144 rem Whole body (immersion) 1.1 rem These doses meet the dose acceptance criteria of 300 rem thyroid and 25 rem whole body as defined in 10 CFR 100.

470.9(RI)-1 W

Westingh0use !

NRC REQUEST FOR ADDITIONAL INFORMATION

.: . ' EE-Response Revision 1 The doses to operators in the main control room were calculated for two modes of operation. In tir one mode of operation, the normal HVAC system (VBS) is assumed to remain in operation with additional air cleanup provided by the supplementary air filtration train. In this mode the MCR is pressunzed by filtered air inflow and recirculation cleanup of the air is provided. The other mode of operation is with the emergency habitability system (VES) operating. With this passive system, there is an inflow of air into the MCR from compressed air storage tanks and there is no air recirculation.

MCR doses (with VBS in operation)

Thyroid 10.9 rem Whole body (immersion) 0.3 rem Skin 8.0 rem MCR doses (with VES in operation)

Thyroid 18,9 rem Whole body (immersion) 0.004 rem Skin 0.2 rem These doses are within the SRP Section 6.4 dose acceptance criteria of 30 rem thyroid. 5 rem whole body, and 30 rem skin.

The above doses were calculated using the following values and assumptions:

Core source data Core activity at shutdown See SSAR Table 15A-3 Release of activity from the core to the containment atmosphere is as desenbed in draft NUREG-1465 with mcdifications described above. Tte percentage of core activity released becomes:

0 - 25 seconds no release at 25 seconds: iodines 0.25 %

noble gases 0.25 %

cesium group 0.25 %

25 sec - 1.0 hr: no release 470.9(R1)-2 W Westingh0use

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 1.0 - 1.5 hr iodmes 4.75 %

noble gases 4.75 %

cesium group 4.75 %

1.5 - 2.8 hr: iodines 35 %

noble gases 95 %

cesium group 25 %

tellurium group 15 %

strontiums 3%

bariums 4%

ruthenium group 0.8%

cerium group 1.0%

lanthanum group 0.2%

>2.8 hr. no release Contamment leakage release data Containment volume (cubic ft) 1.73 E+06 Containment leak rate (% per day) 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.12 1 - 30 days 0.06 Removal of activity from the containment atmosphere Elemental iodme deposition removal coefficient (br-') 1.2 DF limit for elemental iodine removal 200 Removal coefficient for particulates (hr") 0.5 DF limit for particulates removal 1.0 E+03 Primary coolant source data Noble gas concentration Maximum Tech Spec value (See SSAR Table 11.1-2)

Iodine concentration Maximum Tech Spec value (See SSAR Table 15A-1) 470.9(R1)-3 .

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NRC REQUEST FOR ADDITIONAL INFORMATION 1-in =3

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E Response Revision 1 lodme chemical form Assumed to be the same as for iodine in the core 1959 particulate.

4.75% 1 & Hl. and 0.25% organic)

Primary coolant mass (lb) 3.4 E+05 Release of coolant into tie containment The total mass of primary coolant is assumed to be (jected to the containment over 25 seconds.

Containment purge release data Containment purge flow rate (cfm) 8000 Time to isolate purge line (sec) 15 Fraction of reactor coolant iodine that becomes airbome Elemental 0.5 Organic 1.0 Cesium Iodide 0.01 Atmospheric dispersion factors See SSAR Table 15A-5 Offsite breathing rate (m'/hr) 0 - 8 hr 3.47 E-04 8 24 hr 1.75 E-04 24 - 720 hr 2.32 E-04 Main control room (general information)

Volume (ft') 42.260 Normal air intake flow (cfm) 1400 Time assumed for MCR isolation (sec) 15 Atmospheric dispersion factors See SSAR Table 15.A-5 Breathing rate (m'/sec) 3.47 E-04

1. 1. 1. W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION is:

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Response Revision 1 Main control room (VBS in operation)

HVAC volume (ft') 112,000 Filtered air inflow (cfm) 960 Effective indow filter efficiencies Particulates ('7c) 99 Organic and elemental iodines ('7c) 88 Recirculation air flow (cfm) 14,000 Effective recirculation filte efficiencies Particulates (%) 84 Organic and elemental (%) 19 Unfiltered inleakage (cfm) 40 Occupancy factor 0 - 24 hr 1.0 24 - 96 hr 0.6 96 - 720 hr 0.4 Main control room (VES in operation)

Pressurization air flow rate using bottled air (cfm) 20 Filtered air intake (cfm) N/A Recirculation air Gow (cfm) N/A Unfiltered inleakage (cfm) See discussion below Occupancy factor See discussion below The unfiltered inleakage rate and occupancy factor for the MCR are based on the shift schedule and vestibule air volume discussed in Revision 1 of the response to RAI 450.2. For tre first 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of the acciders ingress / egress is discretely modeled with the following bases:

1. Initial shift change occurs at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

2. Shift duration is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months plus a half hour for debriefing period W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1

3. There are three shifts of operators

4. The volume of air in the vestibule is 174 ft 3(taking into account displacement of air by the operators)

5. All members of a shift enter or leave the MCR at the same time to limit entry of unfiltered air into the MCR

6. When a new shift enters the MCR, the inner door is left open (making the outer vestibule door part of the MCR pressure boundary) until the departing shift is ready to exit the MCR.

SSAR Revision: NONE 470.9(R1)-6 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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Ouestion 471.21 Provide updated nonnal operation and post-accident radiation zone maps for the AP600 design. These maps should include information showing all vital areas and routes taken during post-accident operations. This information is needed to complete the responses to the following RAls:

a. Q471.04, which requested that all locations containing radioactive pipe chases be indicated on area maps;

b. Q471.14, which requested that the location of all high and very high radiation areas OIRA/VHRA) be indicated on area maps; and

c. Q471.06, which requested a listing of the number and location of area radiation monitors.

Provide the location of the following on architectural or radiation zone maps:

all radioactive pipe chases and locations within rooms of piping containing radioactive materials; all locked HRA/VHRA; and area radiation monitors (approximate locations)

Note: 10 CFR Part 20 defines a Very High Radiation Area as an area, accessible to individuals, in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in 1 bour at I meter from the radiation source or from any surface that the radiation penetrates. In addition, for purposes of design certification and staff review, the Standard Review Plan requires that areas where personnel could receive 100 rad or more in one hour be identified.

Response

The answer to RAI 471.04 was submitted to the NRC via Westinghouse letter ET-NRC-93-4023/NSRA-APSL-93 0467 dated November 30. 1993. Drawings showing the radioactive pipe chases and location are included in Westinghouse letter ET-NRC-93-4022/NSRA-APSI 93-0464 dated November 29,1993.

The answer to RAI 471.14 was submitted to the NRC via Westinghouse letter ET-NRC-93-4023/NSRA-APSL 0467 dated November 30,1993. A set of drawings showing the access control features is also included with the answer. Drawings showing the updated normal arx! post-accident radiation zone maps and location of high and very high radiation areas were submitted to the NRC via Westinghouse letter ET-NRC-93-4022/NSRA-APSL-93-0464 dated November 29,1993.

The answer to RAI 471.06 was provided to the NRC via Westinghouse letter ET-NRC-93-3936/NSRA-APSL 0297 dated August 9,1993. A list of radiation monitors is included with the answer.

SSAR Revision: NONE 471.21-1 W-Westinghouse i i

NRC REQUEST FOR ADDITIONAL INFORMATION EE EE

= x Question 480.38 When reviewing the AP600 design and igniter placements, some of the enteria that the staff uses are:

a. The optimum placement of igniters is nearest to the source.

b. Compartments adjacent to the break companment should have igniter coverage.

c. Companments that are considered as lower probability break companments should have igniter coverage if tre companments are eccessible.

d. Appropnate test data are the NTS, HCOG 1/4 scale, and HDR programs.

e. Computer analyses, such as MAAP and WGOTHIC, are a valuable tool in assessing general trends.

However, they are not sufficient in determining whether or not igniters are needed in a specific location.

f. Detonation calculations have an important role in the overall assessment of the design. They should be used in two capacities: (1) to show that the compartment can survive the bounding calculations, or (2) to determine the capability of the surrounding stmetures.

g. Placement of igniters should be near the source, but not directly above so as to not be in tie irnmediate buming zone.

h. Equipment survivabdity should be addressed by detennining the environment in the buming zones.

Does Wesungbouse have some similar cnteria that were used to decide upon the placement of the 58 igniters in the AP600 design? Descnbe that criteria.

Response

Chapter 16 of the AP600 PRA and additional details in the response to Question 480.41 provides the Westingbouse enteria for the placement of igniters and the general industry test information utilized.

Equipment survivability is addressed in the response Question 480.48.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION N

Question 480.33 As stated in Q4MO.38, the staff considers the optimum igniter plrement to be near the source. For the AP600, sources include:

. release through stages 1-3 of the ADS into the IRWST

. release through stage 4 of the ADS directly into the containment atmosphere

. release from the high vent points

- release through breaks and seal failures.

Discuss the igniter coverage for each of these locations. How was the placement and number of the igniters decided upon?

Response

The number and placement of igniters is presented in Section 16.2 of the PRA and further discussed in the response to RAI 480.41. In summary, the selection was based on engineering criteria from insights on how the hydrogen released during a severe accident is expected to behave in containment. The primary criteria for an igniter system is to promote hydrogen burning at as low a concentration as possible, and to the extent possible, to burn hydrogen continuously so that the hydrogen concentration will not build up in the containment. To achieve this goal, igniters are placed in major regions of the containment where hydrogen may be released, through which it could flow, or where it may accumulate. The igniter coverage for the specific identified locations are addressed below.

Release thniugh stages 1-3 ADS into the IRWST: Igniters that serve as ignition points for releases into the IRWST include igniters 35. 36,37 and 38 at an elevation of 135 and are depicted in Figure 16-5 of the PRA. The igniters are located at four of the IRWST vents distributed circumferentially around the outer perimeter of the tank and in proximity to the tank vent path during an ADS actuation. Location of the igniters near the tank vents provides ignition if the hydrogen concentration reaches the flammability limit within the IRWST. Locating the units within the IRWST is avoided due to the potential for wetting and cooling the igniter surface.

Release through stage 4 ADS directly into the containment atmosphere: The fourth stage ADS discharges directly into the associated loop compartment at approximately elevation 112'. Igniters that would serve as ignition points for a fourth stage release from reactor coolant loop 1 into loop compartment 01 include igniters 11 and 13 at elevation i15'; 15 and 16 at elevation 125' and 12 and 14 at elevation 130'. For a release from loop 2 into loop compartment 02 coverage is provided by igniters 5 aral 7 at elevation i15: 9 and 10 at elevation 125' and 6 and 8 at elevation 130'. The krations of these igniters is shown on Figures 16-2 and 16-3 of the PR.A which depict the distribution of the igniters around the periphery of the loop compartment.

Release from high vent points: High point vents include a head vent from the reactor vessel head and the ADS valves from the pressurizer. These high point vents discharge to the IRWST through the ADS discharge line.

Therefor igniter coverage for these release points are identical to the first release path discussed aho e.

[ WB5tiligt100SB

NRC REQUEST FOR ADDITIONAL INFORMATION HF :9 L :ii it Release through breaks and seal failures: The most effective oeans of identifying coverage for reactor coolant system is by identifying igniters associated with each compartmen:, room or area. The response to RAI 480.40 provides an identification of primary and secondary igniters which provide coverage for each zone within conutinment. Figures 16-1 through 16-8 of the PRA show the kication of the igniter locations within each zone.

SSAR Revision: None

m. 2 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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Ouestion 480.40 Are there any subcomparti. 'is (such as the reactor cavity) that do not have igniter coverage? If so, what is the justification for not having .o1 mge in those subcompartrnents?

Response

The attached table identifies containment compartments and areas and specifies the hydrogen igniters which provide hydrogen protection for the region. As illustrated each of the containment subcompartments are protected by hydrogen igniters. Although no igniters are physically located within the reactor cavity, protection is provided by igniters in the vicinity that would be exposed to hydrogen from the cavity region. Should hydrogen generation occur in the reactor cavity,large active natural circulation partems would be estabhshed with flow paths along the reactor vessel into the loop compartments and through the reactor coolant drain tank (RCDT) compartment and the open stairwells above the RCDT companment into the upper compartment of the containment. 'Ihe established flow paths provide hydrogen control through mixing as well as numerous igniters located within the path as identified on the attached table. The reactor cavity will normally be flooded following a core damage event. Any igniter located in this region would be flooded and would be ineffective. Also, the need for accessibility for maintenance and testing of the igniters located within the cavity is undesirable from a personnel exposure perspective.11e reactor cavity is protected by igniters 11,13,5,7 and I as indicated in the table.

SSAR Revsion: NONE W Westinehouse o

NRC REQUEST FOR ADDITIONAL INFORMATION Subcompartment/ Area Igniter Coverage SUBCOMPARTMENT IGNITER COVER AGE tELEVATION)

PRIMARY SECONDARY Reactor Cavity II,13, 5, 7 (El I15') 9,10,15,16 (El 125')

1 (El 91') 6, 8,12,14 (El 130')

3,4 (El 95) 2 (El 99')

Loop Companment 01 11, 13 (El 115')

15,16 (El 125')

12,14 (El 130')

Loop Companment 02 5, 7 (El 115')

9,10 (El 125')

6,8 (El 130')

Pressunzer Compartment 50 (El 125')

49 (El 140')

Tunnel connecting Loop 1 (El 91') 30, 31 (El 120')

Compartments 3, 4 (El 95')

2 (El 99')

Soutfrast Valve Room 21 (El 105') 20 (El 105')

Southeast Accumulator Room 20 (El 105') 21 (El 105')

East Valve Room 19 (El 105') 18 (El 105')

Nonheast Accumulator Room 18 (El 105') 17,19 (El 105')

Nonheast Valve Room 17 (El 105') 18 (El 105')

Nonh CVS Equipment Room 33, 34 (El 105') j Lower Companment Area 22,23,24,25 (D 133') )

(CMT and Valve area) 26, 27, 28, 29, 30, 31, 32 (El 120') I l

IRWST Compartment 35,36,37,38 (El 135') .

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480.40-2 W Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION jjElmiEE Subcompartrnent! Area I;: niter Coserage (continued)

SUBCOMPARTMENT IGNITER COVERAGE (ELEVATION)

PRIMARY SECONDARY Upper Compartment Lower Region 39, 40, 41. 42, 43, 44, 45, 46, 47, 48 (El 162')

Mid Region 51. 52. 53. 54 (El 210')

Upper Region 55, 56, 57, 58 (El 235')

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

Hu F g .

Question 480.41 What was the criteria used by Westinghouse to decide on the separation between igniters? On page 6.2-27 of the SS AR it states that the separation between igniter kications is selected to prevent the velocity of a flame front initiated by one igniter from becoming significant before being extinguished by a similar flame front propagating trom another igniter. WCAP-13388 (pg. 4-5) states that the igniters are distributed such that the distance a flame can travel before it encounters another flame front (or exits the compartment) is less than about 25 feet. What is the experimental support for choosing this number?

Response

The selection of the number of hydrogen igniters and their locations for the Containment Hydrogen Control System (VLS) was based on engineering criteria formulated frorn insights on how the hydrogen released during a severe accident is expected to behave in containment. The primary criteria for an igniters system is to promote hydrogen burning at as low a concentration as possible, and to the extent possible. to burn hydrogen more or less continuously so that the hydrogen concentration will not build up in the containment. To achieve this goal, igniters are placed in major regions of the containment where hydrogen may be released, through which it could flow, or where it may accumulate. Among the guidelines utilized in selecting the igniter locations was one based on compartnent geometry and its effect on flame front acceleration. Specifically, in comparments which have long or narrow corridors, a sufficient number of igniters should be installed over the length (or height) of the corridor so that the flame fronts of the igniters need to travel only a limited distance before they merge and thereby limit the potential for significant Game acceleration. Therefore, in compartments with channel-like geometries. the location of igniters in the AP600 containment is based on distributing them such that the distance a llame front can travel before it either encounters another flame front or exits the compartment is less than approximately 25 feet. Additional criteria are provided in Chapter 16 of the AP600 PRA.

The igniter separation selected for AP600 is based on an engineering evaluation that considered the AP6(X) compartment geometries with a background of general industry test infonnation including the following references:

1. Oppenheim, A. K., Introduction to Gas Dynamics of Explosions, Course Lecture Notes, No. 48.

CISM, Udine, Italy, Springer-Verlag, New York, New York, p. 43 (1972).

2. Moen,1. D., Lee, J. H., Hjertager, B. H., Fuhre, K., and Eckoff, R. K., Combustion and Flame, 47, p. 31 (1982).

3. Cousins, E. W. and Cotton, P. E., Chem. Eng.,58, p.133 (1951).

4. Donat, C., Loss Prevention,11, p. 87 (1978).

5. Vehstedt, K. J., Hudson, F. S., and Renfro, D. G., " Combustion Studies at High Hydrogen Concentrations and the Effect of Obstacles on Combustion," Interim Project Report, Whiteshell Nuclear Research Establishment, Manitoba ( April,1982).

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6. Achenbach J. A., Miller, R. B., and Srinivas. V., "Large Scale Hydrogen Burn Equipment Experiments," EPRI Report NP-4354 (December,1985).

7. Ratzel, A. C., " Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests," NUREG Repon CR-4138 (May,1985).

8. Kumar, R. K. Skraba T., and Greig. D., in Proc. of 3rd int'l Conf. on Reactor Thermal Hydraulics, New Port, Rhode Island, p.14.B-1 (October,1985).

9 Kumar, R. K., Canadian J. of Chem. Engng., M p. 866 (1986).

10. Knystautas, R., Lee, J. H. S., Peraldi, O., and Chan, C. K., " Transmission of a Flarne from a Rough to a Smooth-Walled Tube," Progress in Astronautics and Aeronautics, Volume 106, pages 37-52.1986.

I 1. Van Wingerden, C. J. M. and Zeeuwan, " Investigation of the Explosion-Enhancing Properties of a Pipe-rack-Like Obstacle Array " Progress in Astronautics and Aeronautics, Volume 106, pages 53-65, 1986.

12. Sherman. M. P., Tieszen, S. R., Benedick, W. B. and Fisk, J. W., "The Effect of Transverse Venting on Flame Acceleration and Transition to Detonation in a large Channel," Progress in Astronautics and Aeronautics, Volume 106, pages 66-89,1986.

13. Moen.1. O., "The influence of Turbulence on Flame Propagation in Obstacle Environments "

Proceedings of the International Conference on Fuel-Air Explosions, Pages 101-135. Montreal.

November 1981.

14. Lee. J. H. S., Chan. C. K., and Knystautas, R., Hydrogen-Air Dellagrations -- Recent Results."

Proceedings of the Second International Conference on the impact of Hydrogen on the Water Reactor Safety, Pages 889-914, Albuquerque. October 1982.

SSAR Revision: None 480.41-2 W

- WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 480.42 Provide details of the "muhiple electrical power sources" (WCAP-13388, pg. 44) that may power the 58 igniters.

Are they ac-powered or de-powered? What is the pedigree of the power (Class-lE, non-safety, etc.)? Are the igniters divided into any divisions / trains?

If at least some of the igniters are not de-powered, provide a justification for not having de-powered igniters (especially to handle a station blackout).

Response

The hydrogen igniters are non-lE/non safety-related ac powered devices receiving power from the main onsite ac power system (ECS). There is a single group of 58 igniters that receives power from the vital bus of the ECS system. Power is provided to the vital bus from several sources.

Igniter system reliability is provided by the flexibility of the ECS to furnish power to the hydrogen igniters.

The igniter set may be connected to either of the two busses within the main onsite ac power system.

During plant operation the ECS power is supplied by the turbine generator.

When the turbine generator is unavailable (e.g. under accident conditions), the main ac power is provided by the offsite prefened power supply or maintenance source as detailed in SSAR section 8.3.

In the event of the loss of both nonnal and preferred power supplies, onsite standby power to the igniters may be supplied by either of the two onsite standby diesel generators.

The power supply to the hydrogen igniters is not class IE since the igniters are not relied upon to mitigate design basis accidents. Hydrogen control for design basis accidents is discussed in SSAR Section 6.2.4 A single set of igniters provides hydrogen control reliability based on 1) the low probability of core damage events resulting in containment hydrogen concentrations above 109 2) the simplicity and reliability of the igniters and power distribution system tincluding separate protection of igniters potentially kicated in ikxxled regions). 3) the reliability of the ECS to provide power and 4) the small percentage of core melt scenarios without ac power available. The PRA indicates that a very small percentage of the core melt sequences include the loss of both onsite and offsite ac power.

This design approach is consistent with the staWs position on severe accident mitigation specified in SECY-90-016.

SSAR Revision: NONE l

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NRC REQUEST FOR ADDITIONAL INFORMATION

r. - g.g 2r a e

Question 480.43 Has Westinghouse performed any structural response / detonation calculations for compartments in the containment?

For example, if igniters are not included in the reactor cavity, a structural response calculation could show that igniters are not needed there because the structure can withstand the effects of a hydrogen detonation.

Response

Evaluations of the potential for detonations from direct energy deposition and deflagration to detonation transition (DDT) in the AP600 containment are presented in WCAP-13388, "AP600 Phenomenological Evaluation Summaries

The evaluations conclude that detonations in the AP600 containment are highly unlikely. Hydrogen combustion, including detonation, is also being evaluated with decomposition event trees which will be provided as part of Revision 1 of the AP600 PRA. No stmetural assessments specific to detonation in the containment have been performed.

PRA Revision: See above response.

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W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

, . . . w E 21 :

u Question 480.44 Discuss an esent in which there is a large release of hydrogen because of a break in the ADS valve package at the 174 It. elevation. This appears to be at a very high point, with no igniters encountered until the containment dome is reached. What is the potential for large hydrogen concentrations and detonations from such a release?

Response

The ADS valve module, kicated on top of the pressurtzer begins an elevation 158' and continues up to 177'. Two igniters are kicated within proximity of the ADS package; the first is south of the valve module package at an elevation of 162' and the second north of the module at an elevation of 160'. These igniters provide ignition should the hydrogen discharge from the break envelop these kcations. Should the proposed break be higher in the package and be directed upwards, the plume first reaches igniter 53 at elevation 210' and then reaches igniter 55 at elevation 235'. Both of these igniters and to a lesser extent igniter 57 (elevation 235') are in the discharge path above ADS valve package and would serve as an ignition source.

The potential for large hydrogen concentrations and detonation in this region is very low since there are large natural circulation currents driven by the heated releases from the postulated break and the coo!ing xtion of containment shell and there are effectively no barriers for confinement of the break discharges. Additionally, there is no compartmentalization with geometries that increase the potential for detonation.

SSAR Revision: NONE W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION t!E Ein

^

EE Ouestion 480.45 Review igniter perfonnance in the industry today. It appears that the igniters are not very reliable, and that a considerable percentage of them are replaced at each refueling outage when they are surveilled. Given this poor performance / reliability, and, in addition, given that pipe whip or jet impingement may disable an igniter, why is one igniter per room sufficient (as proposed for many of the AP600 subcompartments)?

Response

Single igniters have been placed only in compartments / areas that are either dead-ended with restricted entry flow paths which minimizes the potential for hydrogen inflow and collection. or have open paths to adjacent areas which contain additional igniters. The response to RAI 480.40 delineates the primary igniters and the secondary igniters for containment compartment / areas. The tabulation demonstrates that containment compartments are provided with more than one primary igniter or secondary igniters in adjacent areas are available to backup the primary igniter.

A review of igniter performance in PWR in-containment hydrogen control systems has shown relativiey good experience. Based on discussions with PWR plant users and a review of PWR plant failure repons, the largest ,

number of failures have been attributed to nonnal wearout of the igniter glow plugs due to exceeding the service life.

Other causel categories include commisioning mortality failures of the glow plugs and damage accidentally inflicted by plant personnel during maintenance and inspection activities which contributed to igniter glow plug unit failures.

User experience suggests that both pretesting igniter glow plugs prior to installation and taking care to implement preconditioning practices prior to operation help ensure that the units operate reliably for their normal lifetime. In addition, frequent test monitoring of unit current characteristics supplemented by glow plug temperature surveillance during outage periods detect degradation indicative of approaching the end of service life. A preventive maintenance program implemented by the Combined License applicant to replace igniter glow plug units on a planned basis during normal outage periods considering normal life expectancies for the units further addresses the problem of failure due to normal wear out. The igniter components can be tested periodically to verify operability without affecting the component life or performance. Periodic inspection and testing will be performed to provide ongoing confirmation of igniter operability. See SSAR section 16.2.

To address both the potential for damage accidentally inflicted by plant personnel and also the potential for damage during accident events, the precise location for the igniters will be adjusted after final containment piping and equipment layout is completed. The placement criteria will include the following:

Igniters will be placed to minimize exposure to pipe whip, missiles, or steam water jets.

Gas flow around the igniters will not be impeded by equipment.

+ Igniters will be placed in locations where the potetential for accidental damage by maintenance personnel will be minimized.

W.- WestinEhouse 1

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l NRC REQUEST FOR ADDITIONAL INFORMATION uii 92 1 4 :s The present arrangement of the igniters within containment krates igniters above the lhed level, if possible. Those potentially below the thxd level will be provided with redundant fuses to protect the power supply.

SSAR Revision: NONE l

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NRC REQUEST FOR ADDITIONAL INFORMATION gE =51

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Ouestion 480.47 Has Westinghouse reviewed the following experimental databases? Discuss the application of these programs to the AP600.

The NTS program, which is an excellent source for evaluating the igniter performance within a single compartment.

HCOG, which may provide the most appropriate data for evaluating the performance of transients involving an intact primary system. Most notable of this group is the station blackout sequence.

This group could also be used to gain insights into hydrogen release into the spargers and out into the IRWST.

HDR, which was a test program in which multi-compartmented large scale tests were conducted.

he most significant test is the one selected for the standard problem. This test is of particular interest in that the released hydrogen quickly went to the dome. His test is also good for demonstrating code accuracy. WGOTHIC calculations of HDR hydrogen distCoutions would provide good insights into the design.

Response

ne experimental database (NTS, HCOG, and HDR) programs have been reviewed. He application of these experimental programs to the AP600 design is summarized as follows:

Nevada Test Site (NTS) ne NTS series of tests were conducted in a large (2048 m)) spherica. vessci with mixtures of hydrogen, steam, and air. These tests evaluated igniter performance (burn completeness) in both quiescent and turbulent gas spaces.

He turbulence was generated by an operational fan within the NTS test vessel.

Rese test results were used to conceptualize and benchmark the igniter burn model included in the MAAP 3.0B and M AAP4 computer codes. The model's predicted peak burn pressure, burn completeness and bum duration were compared to the test observations (see Table 480.47-1 (ref. 480.47-1)). A single model parameter (the flame flux multiplier) was varied to assess the values that produced agreement with the test observations for a variety of initial conditions. A value of 2 for the flame flux multiplier for quiescent tests that include a range of both the hydrogen and steam mass fractions was obtained. A value of 10 for the multiplier for turbulent environments is based on the test with fans operable.

He igniter and combustion model have been used to assess the AP600 containment's response to hydrogen bums by performing MAAP4 analyses. He AP600 design does not include containment sprays which would induce turbulence. However, the AP600 design does include a passive containment cooling system (PCCS) which uses an external water spray and natural convection cooling on the outside of the containment shell. The PCCS will cause

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NRC REQUEST FOR ADDITIONAL INFORMATION E!

steam condensation inside the containment on the steel containment shell. He PCCS operation will induce mixing inside the containment gas space which will produce some level of turbulence. This effect is incorporated in the AP600 hydrogen burn assessments by using an intermediate value (~ 5) for the flame flux multiplier. Also, the AP600 design includes non-safety fan-coolers which could promote mixing and increase turbulence. The effect of the fan-coolers is being considered in the e.nalyses for Revision 1 of the level 2 PRA.

The NTS data is also referenced and presented in part in the WCAP-13388 section on hydrogen deflagration and detonation in the AP600 containment. This document discusses steam effects in the AP600 containment gas space on combustion completeness and resultant system pressure rise.

Mark III Hydrogen Control Owners Group (HCOG)

The database produced by the HCOG program is still proprietary and not available in the open literature. It is discussed to a limited extent in the U.S. NRC Safety Evaluation Report (ref. 480.47-2) issued on this tcpic. This document has been reviewed for applicability of the data to the AP600 design.

Diffusive burning on the suppression in the Mark Ill containment was investigated in HCOG's quarter-scale te6t facility. He annular suppression pool, simulated spargers, and structures / floors above the pool where included in the test facility. The facility design features included containment sprays, a top row of LOC A vents, and unit coolers. Hydrogen was released through the submerged spargers and burns were initiated with glow plug igniters.

He principal features of the AP600 design which relate to the HCOG tests are the IRWST and igniters. The IRWST receives discharges from the RPV depressurization system via spargers submerged in the IRWST. This AP600 configuration is somewhat analogous to the suppression pool with submerged safety relief valve discharge spargers employed in Mark III containments and tested in conjunction with igniter performance as part of the HCOG program. The behavior of hydrogen burns in the IRWST gas space is discussed in both the WCAP-13388 section on hydrogen deflagration and detonation and the response to RAI 480.36. Based on the HCOG program results it is expected that the igniters in the IRWST vents will induce hydrogen burns in the IRWST gas space if combustible concentrations are produccd. Due to the submerged discharge of the combustible gases, ?t is expected that diffusion flame on the IRWST pool surface will be produced. The IRWST design is being assessed for such a situation as part of the level 2 PRA Revision 1.

Heissdampfreaktor (HDR) Test Program Large scale hydrogen mixing tests were conducted in Germany in the decommissioned HDR facility. HDR is a 50 m high and 20 m dieneter cylindrical containment with a hemispherical dome and several interconnected rooms in the lower half of the cylinder. The containment is enclosed in a 1.7 cm thick steel shell. There is an air gap ( ~ 0.6 m) between the steel shell and the outer concrete wall. Two of the tests (El1.2 and T31.5) performed in this test program were selected as international standard problems and were used to benchmark the MAAP4 and GOTHIC codes.

HDR test El1.2 simulated a small LOCA by injecting superheated steam for 12 hr to pre-heat the containment.

His was followed by a 30 min injection of a bydrogen/ helium gas mixture and then three additional hours of only 480.47-2 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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steam injection. The containment was next allowed to cool naturally for 15 minutes before water sprays were applied to the outer surface of the containments steel shell. HDR test T31.5 simulated a large LOCA by a 50 second blowdown of a steam and water mixture followed a superheated steam blowdown and injection of a hydrogen / helium gas mixture.

Since the AP600 containment configuration includes an external spray (PCCS) system, test Ell.2 is useful for assessing the mixing inside containment due to natural circulation induced by steam condensation on the steel shell's inside surface. These two sets of test results are also being studied to evaluate stratification potential and the containment's mixing behavior in the absence of igniter operation as part of the level 2 PRA Revision 1.

References:

480.47-1 DOE,1988, Modifications for the Development of the MAAP-DOE Code; Volume III: A Mechanistic Model for Combustion in Integrated Accident Analysis, DOE /ID-lC216, Vol. Ill.

480.47-2 NRC,1990 Safety Evaluation Report Related to Hydrogen Control Owners Group Assessment of Mark III Containments, NUREG-1417.

PRA Revision: See above 480.4F3 W Westinghouse

Table 480.47-1 COMPARISON OF Tile TilEORETICAL RESULTS WITil NEVADA TEST SITE (NTS) PRE 311XED C051BUSTION TESTS Test or Initial Conditions Results Code (CP) Flame Flux Prediction Afultiplier hlaximum Bum ,

Volume Volume Pressure Pressure Combustion Duration Percent 11 2 Percent CO2 (kPa) (LPa) Completeness (sec)

NTSP01 -

5.3 4.2 97.4 144. .32 52.9 CPI i 125. .18 21.4 CP2 2 149. .32 5.2 NTSP03* -

5.8 14.4 95.4 170. .50 12.3 CPI 8 164. .43 1.9 CIU 10 168. .46 1.6 NTSP09 - 6.1 4.2 91.6 167. .60 36.6 CPI 1 153. .43 43.4 CP2 2 183. .52 4.4 NTSPOO -

6.6 4.5 91.6 203. .66 15.2 CPI i 168. .50 45.8 CP2 1.8 195. .57 15.8 CP3 2 214. .65 4.0 NTSP12 -

6.9 28.3 92.6 171. .58 24.

CPI I 147. .35 18.1 CP2 1.8 168. .45 4.5 CP3 2 183. .55 4.3 NTSP13 -

7.8 4.4 97.4 306. 1.00 6.4 CPI I 224. .60 57.0 CP2 1.8 309. .93 3.4 CP3 2 323. 1.00 4.2 NTSP14 -

8.1 38.7 93.5 202. .94 16.8 CPI i 158. .35 10.9 CP2 2 221. .71 4.0 NTSP15 -

9.9 - 4.2 109.3 395. 1.00 3.6 CP1 1 426. 1.00 3.8 CP2 2 428. 1.00 3.2 NTSPl6 - 11.I 27.2 101.3 324. 1.00 4.4 CPI 1 335. 1.00 5.3 CP2 2 339. 1.00 3.8 Igniter located near center of test chamber, Ian used for mixing.

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I NRC REQUEST FOR ADDITIONAL INFORMATION as EE E yy Question 480.48 Discuss equipment survivability and burn zones caused by igniters. Within the burn zone, the environment is likely to eneed the enteria of 10 CFR 50A9. Have all such burn zones been considered, and is safety-related equipment kicated outside of such burn zones?

Response

10 CFR 50A9 requires a qualification pmgram for safety related electrical equipment that is relied upon following design basis events to ensure reactor coolant pressure boundary, reactor shutdown and mitigation of offsite exposures to below 10 CFR 100. A core damage event generating sufficient hydrogen levels to create burn zones is well beyond the defined design basis events and therefore hydrogen burning has not been considered in the qualification of safety related equipment.

As discussed in PRA section 10.2.6, reference 1 states that equipment identified as being useful to mitigate the consequences of severe accidents are to be designed such that there is reasonable assurance that they continue to operate in severe accident environment for the duration they are needed to xcomplish their function. This approach is also similar the requirements of 10 CFR 50.34(f).

The functions of the equipment in the containment, for which credit is taken in the AP600 Probabilistic Risk Assessment study, have been reviewed to determine if the equipment is required to operate in a severe accident environment. The following equipment is credited to function in the PRA after being exposed to a severe accident environment:

+ Electrical and mechanical penetrations

- Hydrogen igniters The design basis for this equipment is to operate satisfactorily in the most limiting severe accident environment.

REFERENCES

1. Attachment to letter from D. M. Crutchfield, Office of Nuclear Reactor Regulation, to E. E. Kintner, Advanced Light Water Reactor Steering Committee, " Major Techrucal and Policy Issues Concerning the Evolutionary and Passive Plant Designs," dated February 27,1992.

SSAR Revision: NONE 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION f

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Respense Revision 1 Ouestion 620.45 What constitutes " review of operating conditions at nuclear power plants" as discussed in Section 18.9.1.1 of the SSAR? Describe how this review was accomplished (p.18.9-1).

Response

Westinghouse's review of operating experience is summarized in the response to Q620.9 referencing the Westinghouse WCAP about such documents as L ERs: the response to Q620,10, discussing ALWR URD requirements; end the response to Q620.12, which discusses interviews with plant personnel. The staffing of the main control room during various plant operations modes will be verified through task analysis and testing.

The first paragraph in Subsection 18.9.1.1 in the SSAR will be revised as follows:

SSAR Revision:

Based on a review of operating experience cond!: en at nuclear plants, two main control area operators are required for plant startup and shutdown. Once steady-state conditions are achieved, only one operator is required for plant operation. -E: c^r rpr::i =y != . O centre! :::= = n=:=rj.

620.45R1-1 W Westingtiouse I

NRC REQUEST FOR ADDITIONAL INFORMATION SrE =ijg Response Revision 1 l

Question 630.3 Section 16.2.3.1 of the SSAR states that the AP600 design organization described in Section 1.4 of the SSAR formulates and implements the AP600 D-RAP.

The staff concludes that the description of the design organization should include the organizational and administrative aspects applicable to the D-RAP, including a discussion on organizational accountability for implementing the design portion of a RAP, and means for disposition of vendor and plant design organization equipment reconunendations. He D-RAP should describe the programmatic interfaces (i.e., how various parts of the design organization interface). The description of the design organization should include how the performance of risk-significant SSCs, when compared to that specified in PRA, will be fed back to the designer to resolve reliability discrepancies.

The staff position is that the D-RAP applies to the certified design applicant and the design entity that completes the site-specific portions of a plant (e.g., an architect / engineer (A/E) under contract or a COL applicant acting as its own A/E). Provide a discussion regarding the D-RAP and its applicability to an architect / engineer in Section 16.2.3.1 of the SSAR.

Response

SSAR Subsection 16.2.3.1 will be revised to provide additional information related to organizational and administrative aspects applicable to the D-RAP.

He AP600 design process involved integration of design activities and PRA risk evaluation of the design. The PRA report and the response to Q720.2 provide an overview of the integrated design and risk evaluation process used for the AP600.

Westinghouse agrees with the NRC staff position that the D-RAP is applicable to the responsible design organization until such time as the O-RAP is in place during plant implementation. This includes the Combined License applicant and the design entity that completes the site-specific portions of the plant.

SSAR Subsection 16.2.3.1 will be revised as follows:

SSAR Revision:

16.2.3.1 The AP600 Design Organization The AP600 design organization described in Section 1.4 formulates and implements the AP600 D-RAP.

The AP600 management staff are responsible for the AP600 design and licensing.

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1 NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 The apt >00 staff coordinates the program activities including those performed within Westinghouse, as well as work completed by the supporting organizations, including architect <ngineers, listed in SSAR Section 1.4.

The AT500 staff is responsible for development of the D-RAP and the lvarious desigr6 analyses, and risk and reliability engineering required to support development of the D-RAP. Westinghouse is responsible for the safety analyses, reliability analyses, and the probabilistic risk assessment.~ The risk and reliability analyses are performed using common data bases pmvided from within Westinghouse and from industry sources such as INPO and EPRI.

The AP600 staff coordinates the system and component design, safety analyses, reliability analyses, and risk analyses to ensure integration.

630.3R1-2

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NRC REQUEST FOR ADDITIONAL. INFORMATION si!" Eli Resposne Revision 1 Question 630.4 Section 16.2.3.3 of the SSAR states that extensive efforts are involved in optimizing the AP600 design for operational availability as well as safety. This section also discusses that the use of consistent reliability information provides confidence that the calculated system availabilities are based on the same data and assumptions as the PRA evaluation. Whenever an alternative design is proposed to improve performance in either area, the revised design is first reviewed to provide confidence that the current assumptions in the other areas are not violated. Additionally, Section 16.2.3.3 of the SSAR states that the identification and prioritization of the various possible failure modes for each component leads to suggestions for failure prevention or mitigation, and that this information is provided as input to the O-RAP because it defines the means by which component reliability can be maintained. He final designs approved for construction reflect the reliability requirements assumed in the design and PRA as part of their procurement specifications.

Although extensive efforts are involved in optimizing the AP600 design for operational availability as well as safety, these ebjectives may, at times, be conflicting (e.g., operational availability goals may be in conflict with the plant safety goals). The staff's position is that it should be clerly stated that safety goals take priority over otber goals whenever a potential conflict exists. Revise Section 16.21 J the SSAR to explicitly state that plant safety goals take priority over other goals.

Response

Westinghouse is committed to meeting the NRC safety goals.

SSAR Subsection 16.2.3.3 will be revised as follows:

SSAR Revision:

16.2.3.3 Design Considerations Extensive efforts are involved in optimizing the AP600 design for operational availability as well as safety. The use of consistent reliability information provides confidence that the calculated system availabilities are based on the same data and assumptions as the PRA evaluation. Whenever an alternative design is proposed to improve performance in either area, the revised design is first reviewed to provide confidence that the current assuinptions in the other areas are not violated. Whenever a potentialconflict exists between safety goals and other goals, these conflicts are resolyed withoutjepapdizing the protection of the health and safety of the public.

As part of the design process, risk-significant components are evaluated to determine their dominant..

630.m a W Westinghouse

1 NRC REQUEST FOR ADDITIONAL INFORMATION EE' En Response Revision 2 Cuestion 720.56 The AP600 PRA does not indicate how the accident management issues discussed by SECY 89-012 will be implemented. Desenbe Westinghouse's planned approach for assunng that each of the tive elements of accident management defined in SECY-89-012 will be appropriately addressed by the vendor and licensee. Identify the respecuve responsibiliues of Westinghouse arxi the licensee for addressing each of the five elements, and any methods and/or guidance that are expected to be used in this process.

Response (Revision 2):

De AP600 plan for addressing the severe accident management prograra requirements discussed in SECY-89-012 is based on the current efforts by Westinghouse on behalf of the Westinghouse Owners Group to develop severe accident management guidance for the current generation of operating plants. From the standpoint of potential severe accident phenomena and potential challenges to the plant fission product boundaries, the AP600 response to severe accidents is bounded by that of the current generation of Westinghouse PWRs. Rus, the ongoing Westinghouse Owners Group program to develop genene severe accident management guidance has direct applications to the development of AP600 plant severe accident management response guidance. It is expected that the respective responsibilities of Westinghouse and the licensee for addressing each of the five elements of SECY-89-012 will be similar to the respective responsibilities of the Westinghouse Owners Group and the licensees for the cunent operating plants. The respective responsibilities are summarized m the following paragraphs.

The framework for the AP600 severe accident management guidance has been developed and documented in WCAP-13913 (Proprietary) and WCAP-13914 (Non-Proprietary). De framework document includes a discussion of severe accident management requirements, the anticipated stmeture for tie decision making process, the goals that must be accomplished for severe accident management, and a summary of possible strategies for AP600 severe accident management. Completion of the development of the severe accident management guidance for the AP600 is part of the man-machine interface development process.

The accident management issues discussed in SECY-89-012 cover a broad range of accident management ac6viues including the symptom-based emergency operating procedures and the utility site emergency plan. The severe accident management issues discussed in SECY-89-012 must interface with both of these. For the AP600, the interface with tir symptom-based emergency operating procedures will be similar to the interface for the current gercration of operating plants (i.e., the transition from emergency operating procedures to severe accident management guidance). While the site emergency plan is expected to be simplified for tie AP600, the interface between the emergency plan and the severe a cident management guidance, on a broad scale,is very similar to that for the current generation of operating plants. The severe accident management guidance must fit the emergency response team responsibilities and authorities, including the chain of commarxi. While generic symptom-based emergency operating guidelines exist to establish a concise interface, the site emergency plan is developed by cach COL applicant, based en specifics of its emergency response organization and interfaces with federal, state and local govemment agencies. Bus, the severe accident management program for the AP600 cannot totally address the issues discussed m SECY-89-012. Issues, such as overall decision-making responsibility and dunes and W Westinghouse

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r Response Revision 2 responsibihties of irxiividuals in the emergency response organizauon and trairung, are interfaces with the COL applicant site emergency plan that can be addressed only in the combined hcense application.

The following is a tugh-level discussion of the method in which Wesungbouse will address each of the severe accident management issues discussed in SECY-89-012 for the AP600:

Accident Management Procedures This element refers to the consideranon of generic accident management strategies identified by the NRC to enhance the abihty to cope with the severe accident scenarios that tend to dominate risk in PRAs for the current generation of operatmg plants. Dese strategies have been identified in several NRC reports, including NUREG/CR-5474 and NUREG/CR-5781. De applicability of the strategies identified in NUREG/CR-5474 for AP600 is discussed in the response to RAI 720.54. De applicability of the strategies identified in NUREG/CR-5781 is part of the insights evaluation discussed in the response to RAI 720.55. As discussed in the responses to RAls 720.54 and 720.55, the applicable NRC strategies are further considered in the development of either generic symptom-based emergency operating procedures or generic severe accident management guidance, as appropnate.

Training for Severe Accidents Training is within the scope of the COL apphcant emergency plan. The specific details of severe accident management training are in the scope of the combined licertse application.

Accident Management Guidance Westirghouse will develop generic severe accident management guidance for the AP600 that provides a means of diagnosing plant conditions during a severe accident and a set of strategies for responding to those plant conditions.

The Westinghouse Owners Group severe accident management guidance, being developed for the current operating plants, will be used as a basis for defining the AP600 severe accident management guidance. From the standpoint of potential severe accident phenomena and challenges to the plant fission product bourxianes, the AP600 severe accident response is bounded by the current gerwration of Westinghouse PWRs. De AP600 severe accident management guidance will incorporate those insights from the AP600 PRA and other apphcable sources, as desenbed in the response to RAI 720.55. The severe accident management guidance developed for the AP600 will provide a means for diagnosing challenges to the plant fission product boundaries, for respondmg to challenges with appropriate strategies, and for returning the plant to a controlled, stable condition. The severe accident management guidance will also identify potential negative impacts (e.g., increased challenge to a fission product boundary) of implementing each of the strategies contained in the guidance. Finally, the guidance will contain information related to the expected plant response after implementation of a particular strategy. The severe accident management guidance will also identify a limited set of computational aids to assist in diagnostics and/or to permit rapid evaluations of the magnitude of some of the negative impacts associated with implementation of a specific strategy.

720.56(R2)-2 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION gF E

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Response Revision 2 Instrumentation The severe accident management guidance relies upon the diagnosis of challenges to fission product boundanes and the diagnosis of a controlled, stable state. Westmgbouse will identify, in the AP600 severe accident management guidance, primary and secondary instrumentation indications for those key parameters needed for diagnosis. This approach is consistent with the approach taken in the Westinghouse Owners Group severe accident management guidelines for current operating plants. Where appropriate, the severe accident management guidano will identify methods for inferring the parameters needed for diagnosis from other instmmentation readings. During the development of the AP600 severe accident management guidance, any insights regarding instrumentation (particularly with regard to instrumentation survivability and readout range) will be documented and further evaluated.

Decision-Making Responsibilities Based on information developed during the Westinghouse Owners Group severe accident management guidance program, the decision-making responsibilities during a severe accident should not change significantly from those already specified in the utility site ernergency plan for existing plants. Tie only significant difference introduced by severe accident management guidance is the broader responsibility for the plant technical support staff to provide recommended actions to the control room staff after core damage has occurred. The tools available to the technical support staff for this broader responsibility are the severe accident management guidance derived from the AP600 generic severe accident management guidelines. Considerations related to decisiordmaking responsibilities during an accident, including severe accidents, are in the scope of the combined license application.

PRA Revision: NONE 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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Response Revision 1 Question 720.70 in Q720.4, Westinghouse was asked to identify areas of the AP600 design that the PRA indicated were important to reducing or maintaining risk, and should be addressed in the ITAACs. Describe how PRA insights were used to develop the ITAACs, DAC, and D-RAP. Include examples. Provide cross references in the PRA between the D-RAP. Technical Specifications, and ITAAC requirements for test and maintenance unavailabilities assumed for all systems, and assumed system availabilities goals (i.e., DAS and DIS). For example, the PRA assumed the unavailability of DAS to be 9.0E-3. Le staff found no unavailability goal for DAS during its review of the D-RAP.

Response

As discussed in the response to RAI 720.4, the AP600 design incorporates insights from several PRA studies. The design information for Tier I (including ITAACs) and Tier 2 (including the SSAR) addresses the appropriate aspects of the AP600 design that the PRA studies indicate are important to reducing overall plant risk.

The PRA provides detailed models of safety-related and nonsafety-related systems that provide mitigation functions for the various initiating events considered in the PRA. It evaluates the event mitigation functions performed by the modeled systems for a specified mission time following the initiation of an event. He various PRA system analyses identify realistic failure mechanisms for the components within a system. To evaluate the system performance, the PRA uses industry component failure data and makes assumptions for individual component unavailability due to testing and maintenance based on data provided in the Advanced Light Water Reactor Requirements Document, Volume III. Appendix A to Chapter 1 ("PRA Key Assumptions and Groundrules," EPRI, Rev. 2, December 1991).

The PRA results provide an evaluation of the system performance against high level safety goals. The results are also used to identify changes that can improve system reliability to perform its event mitigation functions.

He PRA is not used to establish specific availability goals for a system. The PRA is used to evaluate system performance, from the perspective ofits availability and reliability M perform the event mitigation functions. Over i plant life, the component and system unavailabilities will be monitored and tracked using plant data bases as part I of the Operational Reliability Assurance Program (0-RAP) to assure that the high level safety goals, such as core )

damage frequency and large release frequency, continue to be met. l Section 16.2 of the SSAR discusses the D-RAP. The D-RAP is not used to establish specific availability goals for i plant systems. He D-RAP documents the reliability assumptions for AP600 SSCs, including the PRA reliability !

assumptions for the COL applicant's use in developing the 0-RAP. The PRA component m, availability and failure i data assumptions to be included in the D-RAP are documented in Table F-4 of the PRA report. A cross-reference to Table F-4 of the PRA report will be added to SSAR Subsection 16.2.3.4 as indicated in the proposed SSAR revision. Table F-4 of the PRA report includes component unavailability and failure data assumed for the safety-related and nonsafety-related SSCs modeled in the PRA. He D-RAP includes SSC unavailability and failure data assumptions for the systems modeled in the PRA independent of the relative importance for the SSC as determined by the baseline and focused PRAs.

720.70R1-1 W -

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Response Revision 1 As part of the AP600 program for the development of the Tier 1 information and the associated ITAACs, Westinghouse developed screening criteria to establish a logical and consistent basis for determining the need for ITAACs. The screening criteria focus on the safety significance of the SSCs within a system. Each of the AP600 systems were screened against the ITAAC screening criteria.

He ITAAC screening process resulted in developing Tier 1 information and ITAACs for the safety-related systems and for those nonsafety-related systems that provide defense-in-depth functions, independent of the relative importance for the system as determined by the PRA studies.

He AP600 Design ITAAC development process is consistent with the approach used in the development of ITAACs and equivalent information is included in the development of Design ITAACs for a system, structure, or component.

Based on the process used in the ITAAC program, the requirements specified in the ITAACs are independent of the specific testing and maintenance unavailability assumptions in the PRA and therefore cross-references between these documents and the PRA are not appropriate.

He AP600 Technical Specifications were developed by applying the four screening criteria provided in 52 FR 3788 to all AP600 systems. The testing frequencies assumed for specific components modeled in the PRA are consistent with the associated surveillance test interval provided in technical specifications for that component. The surveillance test intervals used in technical specifications can be found in the PRA under the appropriate system analyses provided in Appendix C.

The completion times for action statements in technical specifications are not directly related to testing and maintenance unavailabilities assumed in PRA. The PRA testing and maintenance unavailability selected for a component is based on historical data that considers appropriate planned testing and maintenance performed on the component. The technical specification completion time is determined considering the seriousness and possible consequences of continuing to operate in a degraded condition, the required actions to restore the component and potential consequences of performing those actions, and the time needed to complete the required actions.

Therefore, the testing and maintenance unavailability and technical specification completion times are not directly related and providing a cross-reference between these documents is not appropriate.

He assumed PRA testing and maintenance unavailabilities are not included in either the ITAACs or the SSAR (except for the D-RAP). he assumed PRA testing frequencies are related to certain technical specification surveillance test intervals for a limited number of components. He surveillance test intervals used in technical specifications can be found in the PRA under the appropriate system analyses provided in Appendix C. Therefore, no additional cross-reference beyond that included in the proposed SSAR revision is necessary.

PRA Revision: NONE 720.70R1-2 W WestinEhouse

i NRC REQUEST FOR ADDITIONAL INFORMATIOf; i

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Respomo Revision 1

- 1 SSAR Revision:

Subsection 16.2.3.4 will be modified as follows:

16.2.3.4 Information Available to Combined License Applicant The Combined License applicant is responsible for performance of the O-RAP, which maintains risk-significant SSC reliability throughout plant life. He following information is available to the O-RAP:

The list of risk-significant SSC identified during the design phase, nd t' r :=u n 3 :Ad.446 ni : :! d:: :!:::d z:um;-: c : :!uded i 'h: PR A.

The PRA assumptions for component unavailability and failure data, as provided in Table F-t of the PRA report.

The analyses performed for those components identified to be major contributors to total risk, with the dominant failure modes identified and prioritized. The suggested means for prevention or mitigatior oi 'bese failure modes forms the basis for the plant surveillance, testing and mainte-nance programs.

Table 16.2-2 provides recotamended short-term availability controls : 'N cf d:=gn =c ende-ska for those the-nonsafety-related SSCs ihat perform funtions identified in the RTNSS process.

y d:- Rese recommendations incluoe the operational modes when the systems are risk significant, c:quir-d :c 5: :=i!:b':, +: d:0:2 h dep:F fun:::en: performed by :=F ycen', the recommended modes for extended maintenance operations on the system. and remedial actions if the system is not available.

720.70R1-3 W

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION 55 i=g yi Question 952.12

'Ihe staff has reviewed the design and test matrix for ADS Phase B testing at ENEA Laboratones. It appears that only ore flowmeter is included in the apparatus for measurement of steam Dow. It is recommended that an additional flowmeter be installed on the saturated water lines. Address this comment.

Response

A flowmeter in the saturated water lires wul not be used in the ADS Phase B tests because accurate Dow measurement cannot be obtained in this location due to the two phase flow corvi;tions.

In order to measure the mass flowrate dunng these ADS tests, the supply tr.nk " collapsed liquid" level will be measured using a DP cell. This sunply tank level measurement versus time, with steam / water density corrections based on temperature and pressure wdl provide the supply tank mass versus tirne.

SSAR Revision: NONE W

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION LE' Eii m n-Question 952.13 The effects of non-condensible gases on the operation of the passive safety systems need to be addressed in Westinghouse's test programs for the AP600. The presence of hydrogen in the pressunzer and its introduction into the passive safety injection piping is one example of this concern; a second is the behavior of rutrogen injected into the RCS following injec6on of water from the accumulators. How will these effects be studied in the SPES-2 and OSU test programs?

Response

The effect of noncondersible gases on CMT operation will be investigated as pan of the ChfT separate effects test program. No provisions have been made for simulating the- presence of H2 in the SPES-2 or OSU test facilities.

The introduction of N 2from the accumulators will be simulated in both OSU and SPES-2 test programs since the accumulators wdl contain a scaled volume of N2 (or air), the facility will depressurize at a rate similar to the AP600, and the accumulator delivery flowrate will simulate AP600.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION E

Question 952.14 The staff has previously raised the question of applicability of the passive residual heat removal (PRHR) heat enhanger (HX) testing perfonned at Westinghouse to the current C-tube PRHR design in the AP600.

Westinghouse's responses have concentrated on conditions inside the tubes, and have attempted to demonstrate that the thermal-hydraulic conditions in the test were comparable to those expected during operation of the HX. The sudf has evaluated the test results and Westinghouse's responses to previous RAls; the staff continues to have questions about the ability of the tests to adequately represent the primary side beh:nior of the HX. In addition, the staf f is concerned about the ability to model behavior on the exterior of the PRHR HX tubes, since the C-tube configuration (bundle)is substantially different from the 3-tube configuration tested. 'Ihe staff notes that a C-tube bundle will be included in the OSU facility. However, the relatively short length of this bundle and low operating temperature of the OSU loop will allow only a very limited range of conditions in the PRHR HX to be investigated in a representative geometry; these conditions are not those of greatest interest, such as film boiling and vapor blanketing of the upper portion of the bundle.

a. Provide details of the data base used to predict the heat transfer en the outside of the HX tubes, accounting for the geometry (length, tube diameter, pitch, etc.) and fluid conditions in the IRWST (stratification, subcooled natural convection through pool boiling and possible tube dryout conditions, thennal and vapor plumes, etc.).

b. Aspects of primary side behavior, such as plenum flow distribution and flow distribution among the tubes due to the "C" configuration (with each tube having different pressure drop characteristics and thus different flow) were not studied in the PRHR HX tests. What data base will be used to assist in the modeling of these parameters over the range of expected operating conditions in the AP600?

Response

The PRHR experiments used three full length vertical tubes with a spacing of 1.5 inches between tubes, or a P/D ratio of 2.0. The PRHR tubes were immersed in a simulated IRWST at atmospheric pressure. The IRWST volume / tube was approximately the same as the current C-tube design such that the heat load deposited in the IRWST fluid and resulting boiling and vapor generation effects were similar. Because the IRWST water mass / tube is approximately preserved, the convective flow patterns, buoyancy driven flow, and plume effects are preserved.

Also, the tests were performed over a wide range of IRWST temperatures and primary side Guid temperatures, such that the flow effects on the outside of the PRHR tubes have been captured.

There were specific detailed temperature measurements performed at different elevations to determine the zone of influence that one tube has on the flow. Figures I to 3 show the temperature distribution around the tube and indicates the spacing of adjacent tubes. As the figures indicate, the zone of influence is less than a pitch-to-diameter of 2 which is what was selected for adjacent rod spacing.

W85tingh0USB

NRC REQUEST FOR ADDiflONAL INFORMATION

1 III The current C-tube PRHR rod bundle pitch in the two different directions is shown in Figures 4 and 5. As the figures indicate, the tube-to-tube pitch within a row of tubes is 1.5 inches or P/D of 2.0. The pitch between the adjacent rows is 3. inches or a P/D ratio of 4.0. De spacing chosen indicates, based on the existing PRHR data, there should not be tube-to-tube interactions in the vertical portions of the heat exchanger.

The range of conditions which were examined in the PRHR test program were given in the Response to RAI 440.13 and compared the plant calculation for the C-tube PRRR. As the tables in this response indicated, the range of test conditions covers the calculated plant conditions for the PRHR.

The top and bottom portions of the heat exchangers are horizontal. Comparisons of boiling heat Hux between horizontal and venical surfaces are shown in Figure 6 (Reference 952.14-1). As shown in the figure, for the same wall superheat. the heat Hux on the horizontal surface is higher, therefore, the vertical heat transfer data from the PRHR tests is a conservative estimate of the horizontal tube heat transfer.

The question on possible dryout on the secondary side of the PRHR has been examined. In revie'ving test data, it was observed that the high surface heat Ouxes can only occur when the primary Dow in the tube is high. that is, under conditions that simulate the operation of the main reactor coolant pumps. The highest heas fluxes occur at the top of the heat exchanger where primary fluid temperature is the highest. The calculated surface heat fluxes for these conditions are between 1 x 10' and 2.5 x 10' bru/hr-ft' depending on primary flow and temperature. Dryout was never observed in these cases, and the heat fluxes are below the Zuber pool boiling limit and even further below a flow boiling CHF limit.

When examining the cases which simulate the natural circulation flow on the primary side of the heat , changer, d

the calculated peak heat fluxes are significantly lower,in the range of 2.5 x 10' to 7.0 x I0 btu /hr-ft depending on the primary flow rate and fluid tempemture. Therefore, even when accounting for multi-tube effects, which have been show to decrease CHF in a bundle relative to a tube, the natural circulation conditions result in such low peak Huxes that bundle dryout will not occur under these conditions.

Important factors to PRHR performance are the following:

1) The peak heat ' luxes occur under conditions when the reactor coolant pumps are operating.

2) PRHR performance is most important after the RCPs are tripped. The design basis accident analyses that credit PRHR performance are events w here the CMTs are actuated and the RCPs are tripped. Derefore the the PRHR heat fluxes are lower as described above.

3) Neither the SSAR safety analysis codes nor the PRHR test data predict the PRHR will experience film boiling or dryout, even assuming high flow rates due to the RCPs opemting.

4) The PRHR tube thickness was specifically selected to reduce the likliehood of CHF in the PRHR (the PRHR tube thickness is 0.065 inches compared with SG tubes that have a thickness of 0.W9 inches.)

If dryout did occur in the PRHR under pumped primary flow conditions; the heat load would move down the tubes and nucleate boiling would be estabhshed again and the remainder of the heat exchanger would be more efficient since the primary temperature would be higher. As time progresses, the primary Huid temperature decreases due 952.14-2 W

- WestinEhouse

l NRC REQUEST FOR ADDITIONAL INFORMATION gi: HE!

r y to the heat removed by the PRHR such that at some point. nucleate boiling would be reestablished on all the tube surface area.

The question of the plenum flow distribution and the tube-to-tube flow distribution of the C-tube design has also been examined. The primary flow pressure drop is dominated by the frictional pressure drop in the tubes. The inlet and outlet tube sheet help to establish a more uniform flow in the bundle, but there will be different tube now rates due to the ddferent tube lengths. Calculations indicate that the expected normalized flow per tube ranges from 112%

(for the shortest tube) to 86% (for the longest tube) of the average tube flow rate (assuming single phase Dow). The flow is primarily single phase and the pressure drop is well understood; therefore, the heat exchanger can be modeled successfully as a single average tube, with little uncertainty.

References:

952.14-1 Van Stralen S. J. D. and Sluyter W. M., " Investigations on the Critical Heat Flux of Pure Liquid and Mixtures Under Various Conductions " Int. Journal of Heat and Mass Transfer Vol.12, pp.

1353-1384 (1969).

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

._m W

Wt..:.

FIGURE 1 AP600 PRHR TEST- Plume Test P-02

'230 .

228- -------------------------i----------------------- '

C v ! :

ppg... .. ......................... ............... . ...... ........ ...l.:......... .... .... . . ..... ... ....... . .. . . . . . . . . . . .

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//// .

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-4 -3 -2 -1 0 1 2 3 4 ;

Distance from Tube Centerline (Inches)

-*- TOP -+- MIDDLE -*- BOTTOM 952.14-4 e#

. WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION

= -

.. Y t

l FIGURE 2 AP600 PRHR TEST - Plume Test P-01 230 :

Q....................................... ......................

..........f.... ................ .. ... ,,.. ...... , ,, . ., , . .. ,,,,

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u E 222- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - .

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it 218- ---------------------------i---------- --- -

$216- -- - - - V M ? ~ - '"--- - -- - -- - - -- - -- - -- - - - - -

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v.

210

-4 -3 -2 -1 0 1 2 3 4 Distance from Tube Centerline (Inches)

-=- TOP + MIDDLE x BOTTOM 952.14-5 W WestinEhouse

t NRC REQUEST FOR ADDITIONAL INFORMATION mu 1;*-'

~yt e

E FIGURE 3 AP600 PRHR TEST - Plume. Test C-04 ,

230 .

gg. , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

C V

g m

]Q. ....................................................... . . . . . .

a i F 224- -----------------------i------------ - - -

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212- ----------N -----------~4-------------------- -

v 210 . . . ; . . .

-4 -3 -2 0 1 2 3 4 Distance from Tube Centerline (Inches)

-=- TOP -+- MIDDLE -*- BOTTOM 952.14-6 W-Westinghouse

NRC REQUEST FOR ADD:TIONAL INFORMATION I . . . g

Ci FIGURE 4 PRIIR llX Tube Spacing in llorizontal ilundle a

1.5" d

d Plan View w c1.5"

,T w

3 3"

Section View 952.14-7 W Westinghouse

. ,. ~. _ . ~ ,. . , . . . . .~,

i . ! r- :f l

L NRC REQUEST FOR ADDITIONAL INFORMATION .I i

.j Figure 5 C Tube PRHR Tube Spacing - Full Scale 4

i' i

S .h -

I l L

h

~t ,

M..!

i I- ~9 j

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i g '!

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- i, 952.14-8 i'

[ Westinghouse a + f

NRC REQUEST FOR ADDITIONAL INFORMATION sy: - - -g i

Figure 6 ,

Convection and Boiling IIcat Transfer from Wires at Different Angles (Reference 1) l 1

I 70 - /- .,

Water P1 v6re:N) *0,a t , m so -

10:4, a 5 2, m

. Moeirontal (p = 0) fsy

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IS 20 25 O' S 10 Se. *C 952.14-9

f Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION g,

Question 952.15 Given the experience in ADS Phase A testing of trying to hold the facility pressure constant, how will tests in Phase B be run (e.g., B3-6) so as to hold the pressure constant, as specified in the test matrix?

Response

Constant pressure tests will not be performed during the ADS tests in support of design certification. The test matrix includes only blowdowns through the system with pressure varying over a predetermined range. Testing over a range of pressures produces the data required to perform validation of the safety analysis codes.

SSAR Revision: NONE 952J54 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

sE ;E

= =I Question 952.16 Update the information in the SPES-2 design documentauon to reflect any new hardware or mstrumentation, and to indicate how staf f comments on the test matnx have been taken into considernuort With respect to the test matrix, describe tie key phenornena to be studied in the SPES-2 tests, and justify the choice of these phenomena.

Response

The SPES-2 test facility has been modified to incorporate the following staff comments:

a) The PRHR heat exchanger instrumentation was increased to provide direct heat transfer measurement vs. outside tube wall temperature vs. IRWST water temperature vs. prunary fluid temperature.

b) The PZR to CMT and CL to CMT balance lines were heat traced to prevent excessive heat losses and formation of condensate in the balance lines prior to initiation of the transient. The heat tracing is tumed off when CMT actuation occurs.

The purpose and expected key phenomena for each of the SPES-2 matrix tests is provided below.

Test AP600 Transient Purpose /

Tvte Simulated Key Phenomena / Justificanon Ref. 2-irt diameter cold leg break Reference case for other small break SBLOCA transients - SSAR analysis case.

SBLOCA 2-in. diameter cold leg break with non-safety Investigate system interaction between safety systems operating and non-safety systems.

SBLOCA 2-irt diameter DVI break Direct comparison with reference break (Test No. 3). Investigate asymmetric CMT performance due to increased (subcooled) break flow.

SBLOCA 2-in. diameter break of CUCMT balance Direct comparison with reference break (Test line between balance line isolation valve and No. 3). Initial depressurization from PZR CMT through PZR/CMT balance line until the "S" signal opens CUCMT balance line valve.

Investigate effect on CMT natural circulation beatup/ transition to draining /CMT injection flow.

952.16-1 W-WeStingt100se l

l

NRC REQUEST FOR ADDITIONAL INFORMATION iiii -

SBLOCA l-in. diameter cold leg break with PRilR Provide a significant metease in CMT heatup ilX operating by natural circulation before draindoun begins as compared to the reference break (Test No.

3). Primary temperature reduced by PRIIR operation. Investigate heated CMT behavior durmg ADS.

SBLvCA DEG break of DVI line Complete loss of one of two PXS subsystems.

rninimum PXS injection flow.

SBLOCA DEG break of CIJCMT balance line No delivery from faulted CMT.

between balance hne isolation valve and Depressurization from the PZR, and from the CMT PZR/CMT balance line when the "S" signal opens CL/CMT balance line isolation valve.

No Break Pre-op t*st, inadvenent ADS actuation Confirm facihty depressurization rate / ADS sizing. CMT delivery with ADS apen (no PZR/CMT balance line flow).

SGTR Design basis SGTR (1 tube) with operator Investigate SGTR expected recovery, show actions and use of non-safety systems margin.

SGTR Design basis SGTR (1 tube) with only Investigate recovery and margin to ADS.

passive systems SGTR Multiple SGTR with only passive systems investigate margin to ADS.

MSLB Large steam line break 11vestigate CMT delivery during rapid c3oldown and show margin to ADS.

SSAR Revision: NONE l

l l

">52.16-2 W WestinEhouse

]

?

NRC REQUEST FOR ADDITIONAL INFORMATION

i= ER Ei e g Ouestion 952.17 The following questions refer to infonnation obtained dunng the cold and hot preoperational tesis on tie SPES-2 facihty:

a. Summarize imponant insights on facility behavior that has been gained to this point dunny preopera6onal testing. Include discussion of flow distnbution and pressure losses; facility heat loss charactenstics; instrumentation response and data acquisition; and actual system behavior compared to analytical predictions.

b. Discuss test procedures for matnx testing, with emphasis on any insights gained from preoperational tests.

Response

The requened information will be provided in a pre-operational test report which will be provided by April 30,1994.

SSAR Revision: NONE

[ WB5tingh0USB

NRC REQUEST FOR ADDITIONAL INFORMATION

.ijE %

Ouestion 952.18 Provide the data from the cold and hot preoperational tests on SPES-2.

Response

The requested information will be provided in a pre-operational test repon which will be provided by Apnl 30,1994.

SSAR Revision: NONE 52.18-1 Vj Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

=m .;.

Ouestion 952.19 Provide details on the flow-matclung orifices that will be used in SPES-2 to simulate steam generator outlet and cold leg pressure drops.

Response

Pre-operational testmg (cold, full now operation) at SPES-2 determined that the tiow distribution and pressre drops were acceptable without the insenion of on6ces in the individual cold legs. The results of the preoperatioral tests will be documented in the final test repon armi forwarded to tie NRC at the conclusion of the test program.

SSAR Revision: NONE 4

i M2m J W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

E 'iEi 7

i Question 952.20 Provide a plot and/or table comparing volurne vs. elevation in SPES-2 to that for the AP6(K).

Response

Tius inforrnation will be included in the SPES-2 Facility Decnption Report which wdl be provided by Apnl 30, 1994.

SSAR Revision: NONE

% 2.204

"_N Westinghouse

NRC REQUEST FOR ADDrilONAL INFORMATION j

?k Question 952.21 Provide an analysis to show if there is any difference in plant response to steam generator tube ruptures originating on the cold leg side of the SG, as opposed to SGTRs originating on the hot leg side. -

Response

The SGTR analysis presented in the SSAR modeled a double-ended rupture of a single steam generator tube located at the top of the tube sheet on the outlet (cold leg) side of the steam generator. In response to this RAI an SGTR analysis has been performed medeling the double-ended rupture of a steam generator tube on the top of the tube sheet on the irdet (hot leg) side of the steam generator. All assumptions are identical to those used in the SSAR analysis. The sequence of events and plots of the primary to secondary break flow and ruptured steam generator water volume for the two analyses are compared below. These comparisons demonstrate that the plant response to SGTRs is not altered by the break location.

SSAR Re"ision: None Cold Side Hot Side Break Break Event Time Time tseconds) (se< nnds)

Double ended 50 tube rupture 0 o One themical and volume control system (CVS) pump actuated and pressurizer heater turned on o O Reactor trip on low pressuruct pressure 1074 1146 Main feedwcyr pumps assumed to inp and begm coastdown 1074 1146 Startup feedwater ininated (irriudes maximum delay) 1150 1222 Core makeup tank (CMT) actuation signal on low l pressunzer pressure 1411 1523 Addinonal CVS pump initiated 1412 1525 CMT inpction beg ns (includes maximum delay) 1433 1545 Startup feedwater to faulted SO throttled to maintain low SG narrow range level setpoint 2526 2622 Faulted 50 power operated rehef valve (PORV) fails open when secondary level approumately 3706 3746 reacles high-2 SO narrow range level actpoint Faumd 50 ICRV block valve clones on low steam hne pressure signal (irrtudes valve delay eme) 4346 4440 Sieam release frorn faalied and intact 50 PORVs serminated 4347 4441 CVS and startup feedwater pumps isolated on high-2 SO narrow range level setpomt 5174 5272 Passive residual heat removal system (PRHR) irutisted on high-2 SO narrow range level setpoint 5234 5332 (iruludes valve operung delay tune)

Break flow acrminated and stable condition reached 10 0010 10000 952.21-1 W- WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION Is

0 t

40 b

g 30 CD E

20 s

W CD gg I.

O ' ' ' ' '

~

O 2000 4000 6000 R000 1F+05 12F+05 Time (Seconds)

Comparison of Primary To Secondary Break Flow Following Hot and Cold Side SGTRs 952.21-2 W Westinghouse

I NRC REQUEST FOR ADDITIONAL INFORMATION IEE Eg

n i l

l sat:

- 7000 m

a_

ecco-w r

a

)

h 5000 w - . _

y b

3 4000 O

m a 30C0 w

a t-0- 2CCC a

ex 1000 0

a poca 4aca acao eacc. ,e.cs ,pe.a3 Time (Seconds)

Comparison of Ruptured Stearn Generator Water Volume Following Hot and Cold Side SGTRs 952.21-3 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION is :s Ouestion 952.22 Summarife the important infonnation gamed from the Phase A tests performed on the ADS sparger, including discussion on:

a. straufication behavior in the tank and its impact on the choice of sparger elevation for Phase B testing;

b. tank /intemal loads due to condensation and air cleanng at the sparger, and how these test results apoly to the AP600;

c. experience with water hammer dunng the tests; and

d. unexpected behavior of the sparger dunng the tests.

Response

The specific NRC discussion items are provided below:

a) The operation of the ADS sparger provided energetic mixing of the quench tanL Because of this mixing and the relauvely small temperature increase during ADS operation, temperature strati 6 cation of the IRWST is not expected to occur during ADS. Also, if PRHR llX operation prior to ADS actuation has caused IRWST temperature stratification. ADS sparger operation will mix tie IRWST.

I Temperature stratification had no impact on the choice of sparger elevation.

b) The loads measured throughout the quench tank during the Phase A test have been used to develop a model of the forcing function at the sparger and to confinn the attenuation of the forces as a function of both the distance from the sparger and water depth. This forcing function model was then conservatively applied to the actual IRWST structure to determire the expected IRWST loads and structure responses. See SSAR Appendix 3F for a discussion of AP600 IRWST loads.

l c) No water hammer was experienced during the tests. There was a mild bang experienced during these tests W: 6 steun supply tank isolation valves were unprototypically closed at fast speed l to tern 6 ate the blowdown. This bang was experienced during previous facility test operation and j is due to the limited vacurm breaker size on the facility discharge line. !

d) The sparger did show that choking occurs between the sparger body ard tie sparger arms. This results in three possible choke points in the ADS; the valves, sparger body to anns, and at the j sparger arm holes. Once the second stage valve is fully open. no choking was observed at the I simulated valve locations; but cboking continued in the sparger.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

EHf Another observation frorn the tests is that sparger operation with the quench tank fully heated to saturation produces very low loads on the quench tank but does cause high agitation of the water surface.

SSAR Revision: NONT 952.22-2 W- WestinEhouse

l 1

NRC REQUEST FOR ADDITIONAL INFORMATION l

&=

Question 952.23 I The operation of the passive R11R heat exchangers in the IRWST is expected to have an impact on ADS operation. i especially with regard to changing the temperatum profile of the water in the IRWST before initiation of depressuritation. Were these effects simulated in the Phase A tests? Will they be simulated in Phase B?

Response

See response to RAI 952.22. The mixing caused by the operation of the ADS sparger will eliminate any thennal stratification at tie sparger. The Phase A test did simulate IRWST conditions following PRHR HX operation by pre-heatmg the quench tank water to 180 F and 212 F prior to sparger operation. While there was high agitation of the water surface when the quench tank water was at saturation temperature, no adverse impact on ADS operation was observed. These preheated tests were compared to the tests where the quench tank was initially 90'F.

In the Phase B test, a limited number of blowdowns will be performed with the quench tank pre-heated to 212*F (i.e., maximum possible temperature). These test runs will be performed with the maximum ADS flowTate and energy input into the water.

SSAR Revision: NONT i

l i

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

$F E Question 952.24 Discuss the test procedures to be used dunng Phase B testing of the ADS vahes. includmg

a. duration of tests, and tank temperature profiles before and durmg testing (see Q952.14 on PRHR effects);

b. valve opening times and their relationship to AP600 operation: and

c. measurement of key parameters during tests, such as mass flows, void fractions, and transition from cntical to subsonic flow.

Response

a) The tank temperature profiles are discussed in the responses to NRC Questions 952.23 and 952.22.

The duration of a Phase B test is determined by the amount of steam or saturated water that can be suppbed by the 1300 ft' supply tank versus the ADS stage (s) being simulated. For example, simulating Stages 1,2, and 3 open with a 20 second opening time of the steam isolation valve provided about 20 seconds of blowdown data before the isolation valve began closing in the Phase A test. For the Phase B test runs with saturated water, the expected length of time of the blowdowns is expected to be somewhat less than the steam blowdowns due to the higher mass flowrates that will be achieved. For example, simulation of all three stages of ADS open, with a 10 second isolation valve opening and a 10 second valve closing time should allow ~10 seconds of operation at full flow.

b) The ADS Phase B test simulates each of the three ADS stage flowpaths with full open ADS valves. The initiation and completion of each blowdown test will be performed by facility valves.

The opening times of these valves do not relate to AP600 operation but are determined to supply the desired pressure and fluid conditions for the test. Testing which supports valve selection, including valve opening charactenstics, are not planned as part of the Design Cettification tests.

c) Key test parameters will be measured to enable the determination of mass flows, homogeneous void fraction and transition from/to critical flow. Mass flow will be measured by a venturi during steam flow tests arx! by the supply tank level (mass) change during saturated water / steam blowdown. Multiple pressure / temperature measuwements will provide information on homogeneous void fraction and critical flow locations in the ADS system.

SSAR Revision: NONE 1

i W Westinch00Se o

NRC REQUEST FOR ADDITIONAL INFORMATION l

4 Question 952.25 Desenbe the analyses to be performed using ADS test data from both test phases. Have analyses of Phase A tests been initiated? If so, discuss the analytical results and compare these to the test data.

Response

ADS Phase A The limiting sparger induced pressure pulses measured in the Phase A test were used to determine a pressure pulse function wtuch was tien applied to the AP600 IRWST structure to detennine its hydrodynarnic response and loadings. This information is presented in Appendix 3F of the SSAR (Proprietary Volume 2).

Analyses of fluid flow performance from the ADS Phase A tests have also been performed using the WCOBRA/ TRAC code for the purpose of model development and checkout. One of these analyses used the low flow test data to compute the form loss coefficient of the facility including the sparger and sparger holes. Low flow data was chosen so that there was no choking at the sparger. This information was used to perform simulation calculations of tie test using tte observed form losses.

The form loss coefficient reduced from test A3 is 0.925. Figure 952.25-1 provides a comparison of WCOBRA/ TRAC results and test data. As shown in the figure, reasonably good agreement is obtained for the flow rate, while the predicted pressure distnbution is too high in the large pipe (vils 4 to 13). This over-prediction of the pressure indicates that the flow resistance in cell 3, which consists of .. ace small pipes above the test drum, should be larger than that used in the WCOBRA/ TRAC model for the Phase A tests.

ADS Phase B The Phase B portion of the ADS test will include both steam only blowdowns and blowdowns with a mixture of steam and saturated water. The facility ir cludes a simulation of the AP600 piping atxi valve package with both minimum and maximum valve flow areas and resistances.. Analyses of these tests will be performed with both tl e WCOBRA/ TRAC and NOTRUMP codes.

SSAR Revision: NOST

[ W85tingh00S0

l NRC REQUEST FOR ADDITIONAL INFORMATION

-- n

?? -

t'

' WCOBRAffRAC MODEL.1 ADS Phase A Test Test A.1 80 seconds SPARGER TEST DRUM Ikidy. Arm ' ilotes Valve Ww - - ~

r .

. . . m: llBREAK +

t 0, -

- ; : 3) 3

- i w . ___ .;

2 3 4 5 . ..

13 ' 14 15 16 17- 18' Celi 1 Test Data (psia): P=67.2 30.7 31.5 30.7 32.8 26.8 23.8 -

ECOBRA/ TRAC: P=67.1 44.6 43.8 . . . 40.5 40.1 33.2 29.2 24.4 31.4 23.0 23 8 ,

Test Datas Flow Rate = 84.4 lbui/sec 4 MCOBRA/ TRAC Flow Rate = 80.8 lba/sec )

r Note: 1. Since the ADS Phase A Tests are conducted with steam only, the elevation head is -

negligible. Derefore, in the ECOBRA/ TRAC model all pipes are mmlelled as horimntal pipes. :} .

2.' Although the length of the sparger arm is 4 ft the average from the sparger Iwnly to the sparger holes is 2 ft (cell 18).

3. In the above ECOBRA/ TRAC mmici, the valve is not to simulate the ADS valve.' It .

i is there just to make it easier to reach steady state. He valve is initially closed, it e

ramp to fully open in 5 seconds so that the steam now will reach full flow in a smooth manner..

4. The BREAK component I is set to the drum pressure of 67.2 psia, and the BREAK ' ,

compinent 3 is set to 23 8 psia which is equal to the atmospheric pressure plus the -

water head above the sparger arms in the quench tank. .

5. The sizes of the pipes in the ateve model are appresimately propirtional to the actual . .!

sizes of the pipes, while the lengths are mit.

Figure 952.25 l' .q Comparison of WCOHRAfrRAC Results and ADS Phase A Test Data .

t

-952.25-2 a W Westinghouse .

l

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T a- py , _ ,. ,

NRC REQUEST FOR ADDITIONAL INFORMATION Question 952.26 Provide the following informanon concerning the ADS tests:

a. Ilow will the fluid entering the depressurization valves be conditioned before reaching the valves?

Tlus includes single-phase liquid, two-phase mixtures, and single-phase steam.

b. What is the basts for choosing the specified cor'.utions?

c. Wdl the fluid experience a range of thennal-hydraulic corxtitions similar to that expected in the AP600 dunng the course of the test program? If appropriate, refer to any pretest analyses performed on the test facility or on the AP600 design.

Response

a. The test facihty supply tank pressure, temperature and initial level can be controlled. The open position of the 12-inch facility control valve can also be controlled. The tank pressure, temperature and initial level will determine the duration of the transient. The position of the 12-inch control valve, wtuch is upstream of the ADS valve package, will control the pressure drop and the amount of flashing that can occur across this valve.

b. The basis for choosing the specific conditions for the ADS Phase B tests will be the calculated AP600 fluid corxiitions (pressure, mass flow, temperature and quality) at the valve locations.

c. By choosing different tank pressures, temperatures, and initial levels and by choosing the position (and resulting pressure drop) for the 12-inch control valve, the mass flow and quality at the ADS valve package can be varied such that the tests will produce data over the range of fluid conditions expected for the AP600 as calculated by the NOTRUMP code.

SSAR Revision: NONE l 1

l l

I W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION in? E5

n. e-Ouestion 952.27 The stall is concerned that the full range of possible conditions under which ADS may operate has not yet been considered. Justify your selection of operating conditions. A " map" of RCS behavior of the AP(G) vs. ADS test f acility operation would be useful in making this assessment.

Response

The requested map of RCS behavior of the APful vs. ADS test facility operation will be provided in a revision to this RAI by March 31,1994.

SSAR Revision: NONE WB5tingh0USB

NRC REQUEST FOR ADDITIONALINFORMATION n t:.

iii Question 952.28 Provide information on dynamic forces associated with reflection waves off ridged surfaces during ADS operation.

Response

Analyses of the ADS tests are described in SSAR Appendix 3F. The test tank was a reinforced concrete circular tank which did not have ridged surfaces. Appendix 3F demonstrates that the finite element analyses using the source pressure method are capable of predicting the pressure responses at the tank boundary. This methodology has been used in prior hydrodynamic analyses. Additional parametric analyses have been performeu to investigate the effect of the tank boundary characteristics as described below.

Reflection waves off the tank boundary are dependent on the dynamic characteristics of the tank boundary. A series of parametric analyses of the test tank tank configuration have been performed.Dese analyses investigated the effect of different tank wall flexibility as follows:

o Base case using stiffness characteristics of the test tank o Flexible case having wall frequency of about 9 bertz similar to the outer steel wall of the IRWST o Intermediate case having wall frequency of about 30 bertz similar to the inner concrete wall of the IRWST Five different source term pressure time histories were investigated. These were selected from the ADS test results to have input frequencies in the range of the parametric test tank models. Pressures calculated in these parametric analyses were evaluated at various locations in the tank with the following observations:

o The pressure response close to the source retains the same frequency characteristics as the excitation and is not significantly affected by the tank wall conditions. Pressure amplitudes are generally attenuated as the distance from the sourn increases.

o The pressure response at the water wall interface is affected by the wall characteristics. Both the frequency content and amplitudes are affected.

These additional parametric analyses have confirmed that the pressure amplitudes and frequency content close to the source is only slightly affected by the tank wall characteristics. This supports the use of the source term methodolgy described in Appendix 3F for evaluation of the IRWST for hydrodynamic loads.

SSAR/PRA REVISION: NONE W W85tillEh0US8

NRC REQUEST FOR ADDITIONAL INFORMATION 8 Ii!

Question 952.29 Provide a'l the bend / elbow radius in the following piping systems of the SPES-2 faciht) tall page numters refer to Dwg. 0(T189DD92):

a. Downcomer-Upper liead bypass (pg. 30): 2 elbows

b. Pressurizer to CMT balance lines (pg. 31): 1I elbows

c. Cold Leg to CMT balance lines (pg. 32): 13 elbows

d. CMT Discharge lines (pg. 33): 10 elbows 4 bends

e. IRWST Injection lines (pg. 34): 14 elbows

f. PRHR HX, Supply line, Retum hne (pg. 35): 7 elbows

g. Accumulator Injection lines (pg. 36): S bends or elbows

Response

The SPES-2 Facility Description Report will contam updated (as-budt) piping drawings and will include infomiation about each piping run. This document will be provided by Apnl 30,1994. The following infonnanon is provided in response to the above request:

a. Upper head bypass elbows are 1 inch diameter,long radius ANSI elbows. r=38.1 mm

b. Pressurizer to CMT balance line elbows are bends with 70 to 80 mm radii.

c. Cold leg to CMT balance line elbows are bends with 110 mm radii.

d. CMT discharge line elbows are bends with 110 mm radii.

c. IRWST injection line elbows are bends with 85 mm radii.

f. PRHR heat exchanger supply and retum line elbows:

- 57.15 mm radii for 1.5 inch pipe

- 28.57 mm radii for 0.75 inch pipe 110 nun radii for berxis (no dots)

g. Accumulator injection line elbows are bends with 130 to 140 mm radii.

1 SSAR Revision: NONE l

W Westinghouse l

NRC REQUEST FOR ADDITIONAL INFORMATION i =s Ouestion 952.30 Provide the type and size of valves in the following piping systems of the SPES-2 facihty (all page numbers refer to Dwg. 00189DD92):

a. Downcomer-Upper Head bypass tpg. 30):

1. Isolation valve at Elevation +1425

b. Pressurizer to CMT balance lines (pg. 31):

1. Check valve in piping run from Elevation +10714 to Elevation +10528 (to CMT-A).

2. Check valve in piping run from Elevation +10911 to Elevation +10734 (to CMT-B).

c. Cold Leg to CMT balance lines (pg. 32):

1. Air operated valve in piping run from Elevation +5833 to Elevation +5970 (to CMT-B).

2. Air operated valve in piping run from Elevation +5881 to Elevadon +6027 (to CMT-A).

d. CMT Discharge lines (pg. 33):

1. Motor operated valve in piping run at Elevation -791 (from CMT-A).

2. Motor operated valve in piping run at Bevation -791 (from CMT-B).

c. IRWST Injection lines (pg. 34):

1. Isolation valves just below the IRWST (both lines).

2. Check valves in piping run from Bevation -991 to Elevation - 508 (both lines).

f. PRHR HX, Supply line, Return line (pg. 35):

1. Air operated va've in piping run at Bevation +8026 (Supply line).

2. Motor operated valve in piping run at Bevation +2667 (Return line).

g. Accumulator injection lines (pg. 36):

1. Two air operated valves in piping run from Elevation -3569 to Elevation -508 (one in each accumulator injection line).

W Westinghouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION

Response

The SPES-2 Facility Description Report will contain a complete listing of valve sizes and types. 'Ihis document will be provided by April 30,1994.

SSAR Revision: NONE 95230 2 W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION i=.E 5 Ouestion 952.31 Provide updated and complete schernatics/ plant drawings for the following systems of the SPES.2 facility. The exact locatmns and lengths of the valves or flow meters in these systems should be clearly visible in the schematics.

a. PRI-IR Supply, Heat Exchanger, and Rerum lines.

b. ADS Stage 1,2,3, and 4 inlet, valve nest, and discharge piping lines.

c. CMT Discharge line.

d. IRWST Injection line.

e. Steam Generators in particular the separator bypass.

Response

The SPES-2 Facihty Desenption Report will contain updated (as-built) piping dnwings and will include lengths to each component. This document will be provided by April 30,1994.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 952.32 Clanfy/specify the following operational parameters or particular component operation:

a. Is the leakage power of 1.3 percent from the power rod cable mns through the lower plenum a portion of the total power of 4.9 MW or additive to the total power?

b. Provide the steam generator's steady-state secondary water levels, secondary water mass, recirculation ratios, and riser inlet temperatures.

c. Provide the pressurizer's steady-state water level, the power inputted into the pressurizer using the extemal heaters, and tic control logic for these external heaters.

d. Provide the initial PRHR valve positions and operations dunng tests involving the PRHR system.

e. Provide the ADS valves' rated flows and pressure drops, and for a given flow, tte pressure drops from the ADS valves to their respective catch tanks.

f. Provide the expected heat transfer rate from the CMTs to their outer tank. In particular, the pressure and temperature of the air surrounding the CMTs and what control systems will be used to control the temperature and pressure in this outer tank.

g. Provide the expected volume and fluid temperature of tic Lower Plenum that will become stagnant at tSe sery bottom of tre pressure vessel.

Response

The following information relates to the SPES-2 test facility:

a) The power loss from the heated rods into the lower plenum portion of the power channel is part of tte total power (Note: the current test procedure specifies that the total heated rod power shall be 4.99 +0/-0.1 MW; where 4.99 MW corresponds to 102% of the scaled AP600 rated thermal power.)

b) This infomtation is based on SPES-2 (AP600) operating conditions and will be provided after test dsta on these conditions becomes available.

c) The pressurizer steady state water level is 3.78

0.38 meters as measured by L-010P.

The PZR external heater total power is 22.8 kW (3.8 kW for each of the six heaters) Two beaters are connected to each of tluee on/off power controllers. The heat input adjustment is accomplished by selecting the on/off time fraction.

952.32-1

.N Westinghouse

llRC REQUEST FOR ADDITIONAL INFORMATION i

Use of the extemal heaters dunng steady iruttal condition operation or during the matnx tests is not planned.

d) The PRHR heat exchanger retum line isolation will be closed prior to transient initiation. This valve is opened when either SG's narrow range level is at 5.6% (1 1205/L-B20S = 0.492 ft.) plus a delay time of 45 seconds. This valve is also opened when ADS stage 1 is actuated (either OIT volume is at 75% - L-A40E/L-B40E = 15 ft.). This valve would then remain open for the remainder of the test.

e) The flow through the ADS stages 1. 2 and 3 is detennined by orifices which are 1/395th of the flow area of the two AP600 ADS stage valves represented. e.g.:

AP600 ADS SPES-2 ADS Valve Area Orifice Area ADS Stage 1 2 x 6 in: 0.03 in:

ADS Stage 2 2 x 28 in' O.14 in' ADS Stage 3 2 .t 28 in' O.14 in' Each of the ADS stage 4 flow paths contains an orifice that is 0.52 in'. His erifice sizing has 2

been based on pre-test analyses and is >l/395th of the 75 in AP600 stage 4 valve flow area.

The actual SPES-2 ADS valves and piping are oversized as compared to the AP600 and have a relatively small dP.

An orifice may be installed downstream of the ADS. Stage 1, 2, and 3 turbine meter and y-densitometer to maintain the volumetric flow within the turbine meter range. The final piping dP vs. flow will be provided in the Facility Description Report. His report will be prosided on or before April 30,1994 f) The G1T external tank is pressurized with air to -900 psig. This pressure is maintained by a simple air pressure controller. A PORV set at 972 psig is used to relieve air pressure when the air is heated during G1T operation. No temperature control is provided.

The heat transfer rate from the GiTs to their outer tank and surrounding air will be measured during hot pre-operational tests and will be included in the teduced data.

g) The volume of the power channel lower plenum is 1.13 ft' and has been designed to correspond to 1/395th of the AP600 reactor vessellower head volume. The temperature of the lower plenum l observed dunng simulated full power operation was $41 F . l l

l SSAR Revision: NONE l 952.32-2 W Westingh0use

NRC REQUEST FOR ADDITIONAL INFORMATION HHi .Hj+

=;

Ouestion 952.33 Clarify the position relationship of the pressure vessel connections of DVI A and B with respect to the pressure vessel cold leg nozzle attachrnent locations. Is DVI A between Cold Legs A2 and B1 or between Cold Legs B2 and Al?

Response

The SPES-2 loop piping designations are as follows:

Loop A is the loop containing the PZR and PRHR HX connections. This loop therefore has hot leg "A" and cold legs "Al" and "A2" Loop B is the loop to which the CL to CMT balance lines are connected. This loop has hot leg "B" and cold legs "B1" and "B2".

The reactor vessel nozzle order starting with hot leg "A" at 12 o' clock and going clockwise is: cold leg "Al", DVI "A", cold leg "Bl", hot leg "B", cold leg "B2", DVI "B", and cold leg "A2" The above SPES-2 nozzle designations are for the SPES-2 facility only and do not correspond with the current AP600 plant nozzle designations.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

..,u f ~ qg Question 952.34 Provide the following information on the SPES 2 test procedures.

a. The location, size, and geometry for all breaks.

b. Which ADS 4th stage valve will be failed?

c. Which CMT will be failed?

Response

a) The locations, size, and geometry of the breaks are as follows.

Test Break SPES-2 Break No. Simulated Break Size Position 1.2 1" diam. CL break .050" diam. Bottom of CL"B2",just upstream of the CL to CMT balance line connection.

3,4 2" diam. CL break .101" diam. Same as Tests I and 2 above.

5 2" diam. DVI line break .101" diam. Bottom of DVI line "B",

between CMT/ACC/lRWST check valves and the RV downcomer.

6 DEG break of DVI line .343" diam., two pipe DVI line "B", between (6.813" diam.) ends check valves and RV downcomer.

7 2" diam. break of .101" diam. On the CMT side of the CMT B CL to CMT CL to CMT balance lire balance line isolation valve (balance line off CL "B2").

8 DEG break of a CL to .343" diam., two pipe One break on each side CMT balance line ends of the balance line (6.813" diam.) isolation valve (balance line off CL "B2").

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION dE

ij e

9,l0 Complete break of one SPES-2 break u til Break is from the SG tube. 607" diameter simulate 1.2 X's the tube pump "B" suction to the area secondary side of SG "B" 11 Complete break of 3 SG Break is from the tubes pump "B" suction to the secondary side of SG "B" b) The ADS Stage 4 that is connected to Loop B will be the single failure simulated in the tests (where applicable).

c) There is no single failure which can prevent the operation of a CMT However, tests 5,6,7, and 8 are breaks which will affect the operanon of the affected CMT. In each of these tests, the break location will affect CMT "B" which is connected to CL "B2" and DVI hne "B".

SSAR Revision: NONE 952.34-2 W Westinghouse

e NRC REQUEST FOR ADDITIONAL INFORMATION SE !!Q Question 352.35 Provide infonnauon on the insulation material, thickness, and lengths for the entire SPES-2 facility. This request is in addition to previous inquiries on the measured heat losses of the SPES-2 facility (Letter from N. L Liparulo, Westinghouse, to R. W. Borchardt. NRC, dated August 13, 1993).

Response

l The SPES-2 Facility Description Report will contain information on the insulation material used and thicknesses.

This document will be provided by Apn130,1994. The following infonnation is provided in response the above request:

LOCATION ROCK WOOL THICKNESS (mm)

Power Channel lower plenum bottom 120 lower plenum upper side 100

- core (not insulated) upper plenum / upper head 100 Primary Pump 100 l Pressurizer 110 Steam Generator pnmary plena 120 riser vessel 100 tubular downcomer 70 upper vessel 100 steam dome 80 liot legs 100 Cold legs 100 Power Channel downcomer 100 Power O annel core bypass 100 Power Channel downcomer bypass 70 Pressurizer surge line 80 Main steam line 90 Main steam beader 90 Steam dump 90 Feedwater header 70 Feedwater lines 70 Preheater lines 70 SSAR Revision: NONE W Westinghouse 1

E __

i

1 I

l NRC REQUEST FOR ADDITIONAL INFORMATION ME EE i

1 Question 952.36 l l

Page 8 of Dwg. 00189DD92 shows a cross-section of the upper plenum and annular downcomer in the SPES-2 I f acihty. The 8 circumferential pressure drop fins and the 2 hot leg baffle plates can be seen. Ilowever, there are 4 '

l other structures shown in this cross-sectional view that correspond to the same vessel azimuthal angles as the cold leg nonjes. What are these devices, their physical dimensions (including elevauon in the annular downcomer), and ,

their purpose in the SPES-2 facility, including their relatiortship to any feature or expected phenomena in AP600's downcomer or vessel?

Response

The four " devices" shown in the annular downcomer at the same angles as tie cold leg nozzles (Page 8 of Doc. No. 00189DD92) are lugs which center the annulus. These lugs are shown in the elevation view at the bottom of the annulus,just below the nozzle which connects to the tubular downcomer. They have no impact on tre annular flow patit SSAR Revision: NONE

[ W85tingh0USB

NRC-94-4059, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 from Ltrs of 931027,1102 & 30.Listing of NRC Requests for Addl Info Responded to in Ltr Contained in Attachment a

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SUBJECT:

Dear Mr. Borchardt:

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