ET-NRC-92-3777, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 from 920923 & 1001 Ltrs.Transmittal Partial Response to Ltrs.Listing of NRC Requests for Addl Info Responded to in Ltr Contained in Attachment a

Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600 from 920923 & 1001 Ltrs.Transmittal Partial Response to Ltrs.Listing of NRC Requests for Addl Info Responded to in Ltr Contained in Attachment a

ML20128A141
Person / Time
Site:	05200003
Issue date:	11/30/1992
From:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
ET-NRC-92-3777, NUDOCS 9212030207
Download: ML20128A141 (166)

Text

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c Westinghouse Energy Systems g3;5g ,,,,,,333333 Electric Corporation ET NRC.92-3777 NSRA APSI 92-0257 Docket No.: S't%52-003 '

0'd' N November 30,1992 Document Control Desk U.S. Nuclear Regulatory Commission Washington. D.C. 20555 ATTENTION- D6. THOMAS MURLEY

SUBJECT:

WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600

Dear Dr. Murley:

Enclosed are three copies of the Westinghouse responses to NRC requests for additional information on the AP600 from your letters of September 23,1992 and October 1,1992. This transmittal is a partial response to those letters. A ILsting of the NRC requests for additional information responded to in tais letter is contained in Attachment A. The Westinghouse responses to the remainder of the requests for additional inforrnation contained in yovr letters of Septembet 23,1992 and October 1, 1992 will be provided prior to January 23,1993.

If you have any questions on this material, ;1 ease contact Mr. Brian A. McIntyre at 412-374-4334.

Nichol s J. r Manager Nuclear Safety &yl Acuvities Regulatory

/nja E' m ore cc: B. A. McIntyre - Westinghouse F. Hasselberg - NRR p _

03000- \

1 3

921203020/ 921130 FDR- ADOCK 05200002 I A PDR ow2A

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ATTACHMENT A

- RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC-92-3777)

RAI Number Issue 100.2 NEPA/SAMDA 100.4 Acronyms / Definitions 210.2 Relief and Safety Valve Testing 210.12 Modal Piping Analyses 210.13 Piping Analyses 210.14 Piping Analyses 220.1 ASME Design Report 220.2 Containment loading analysis 220.3 Containment Shear and Tension Connectors 220.5 Con.tainment Ultimate Capacity 220.6 BOSOR-5 Analysis 220.7 Containment Yield 220.13 Equipment Hatch Buckling Factor 220.14 Equipment Match Seal .

220.17 Dynamic Sn Bearing Capacity -

220.18 Section 3A.3.1.3 Equations 220.19 Allowable Stresses 220.20 Concrete Reinforcement Connections 230.7 Component Seismic Damping Values 230.8 Composite Seismic Damping Model 230.14 Stick Model Truss Elements j 230,16 ' Axisymmetric Containment Shell Model 230,19 Enveloped Floor Response Spectra 230.20 Earthquake Motion Analyses 230.21 Response Spectrum Analyses of Structures 231.2 Seismic Floor Response Spectra 1

A'ITACIIMENT A RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC-92-3777)

RAI Numhet issue 231.4 Minimum Soil Bearing Strength 231.6 Lateral Earth Pressure leads 231.7 Stability Analyses 231.8 Soil Liquification 231.9 Seismic Margin Analyses 250.1 Con'ainment Vessel ISI 250.2 Containment Vessel Corrosion ISI 250.3 Component ISI Capabilities 250.4 Component Preservice Inspection 250.5 Preservice inspection Equip nent 250.6 Component ISI Accessibility 250.7 Changes to Section XI Requirements 250.8 Piping Erosion / Corrosion 250.22 Class 2 & 3 Component ISI Access 256.23 Preservice Inspection Requirements 250.24 Preservice inspection Equipment 250.25 ASME Section XI Accessibility 250.26 NDE Requirement Changes 250.27 Piping Erosion / Corrosion 250.28 Turbine Rotor Preservice Inspection 250.29 Turbine ISI Program 251.1 Turbine Maintenance / Inspection Program 251.24 Turbine Rotor Flaws 251.25 Turbine Rotor Fracture Transition Temp 251.26 Turbine Rotor Charpy V-Notch Energy 251.27 Turbine Rotor Fracture Toughness 251.28 Turbine Rotor Brittle Fracture 2

ATTACIIMENT A RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC-92-3777)

I RAI Number Issue 251.30 Turbine Rotor Central Bone 251.31 Turbine Rotor Inspection 252.20 Containment Vessel Materials 252.36 CRDM Transients 252.40 CRDM Use ofinconel 750 252.42 Possible revision to section title ,

252.57 RCPB Fracture Toughness Properties 252.91 RV Material Surveillance Program 252.97 Replacement Sury Capsule Installation 252.98 IAcate Table 5.3-7 252.118 ESF materials list 252.119 ESF materials list 252.126 Hydrogen generation from the corrosion of materials within containment 252.134 Demineralized water makeup system materials 252.135 Failure of the demineralized water makeup system does not jeopardize performance of systems required for safe plant shatdown 252.140 Erosion / corrosion of the steam and feedwater system 252.144 Condensate polishing system radiation contamination cleanup precautions 260.1 initial Test Program 260.2 Initial Test Program 260.3 Initial Test Program 260.4 Initial Test Program 260.5 Initial Test Program 260.6 Initial Test Program 260.7 Initial Test Program 260.8 Initial Test Program 3

4

A1TACitMENT A RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC 92 3777)

RAI Number Issue 260.9 initial Test Program 260.10 Initial Test Program 270.1 Equipment Qualification 270.2 Equipment Qualification 270.3 Equipment Qualification 281.1 RCS Water Chemistry Specification, 281.10 Demineralized water makeup specifications for halogens and sulfate 281.14 CVCS makeup pump capability 281.15 CVCS hydrotest pump 281.16 Mixed bed and cation bed dem'aeralizers 281.17 Safety precautions for storing hydrogen 281.18 Safety Analysis credit for CVCS makeup 281.19 CVCS Ilydrogen supply line isolation 410.2 External Ekmding 410.10 IRWST Lines 420.1 EMI effects on digital I&C 420.3 PMS Process Illock Diagrams 435.2 Three-tier ac system 435.4 Spare unit auxiliary transformer 435.6 Periodic inspection and testing of offsite power system 435.11 RCP trip safety function basis 435.14 Grid stability analysis 435.15 Plant operating limits (real and reactive power, voltage, frequency, etc.)

435.18 Cable derating based on passing through fire barriers 435.19 Load sizing of Class IE batteries 435.20 Load sizing of Class IE batteries 4

A'ITACIIMENT A RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC-92-3777)

RAI Number issue 435.21 Class IE battery load definition 435.22 Class 1E battery load definition (all loads necessary for post-accident monitoring and plant control following licensing basis events) 435.23 Battery charger capabilities 435.60 Containment electrical penetrations 440.1 SPES test, Check Valve testing and PRHR testing 440,4 WCAP 13345 (CMT Testing) 440.5 WCAP 13345 (CMT Testing) 440.6 WCAP 13345 (CMT Testing) 440.7 WCAP 13345 (CMT Testing) r 440.8 WCAP 13345 (CMT Testing) 440.9 WCAP 13345 (CMT Testing) 440.10 WCAP 13345 (CMT Testing) 440.11 WCAP 13342 (ADS Testing) 440.12 WCAP 13342 (ADS Testing) 440.15 WCAP 13342 (ADS Testing) 440.19 WCAP 13342 (ADS Testing) 450.4 Chlorine release _

450.5 Location of plant vents -

450.7 ESF atmospheric cleanup 460.2 SG blowdown rates 460.7 Radiation Monitoring 620.6 Man-Machine Interface system definition 620.7 M-MIS definitions 620.11 M-MIS generalliterature 620.18 "near full-scope, hi fidelity simulator" _.

620.22 Acceptance criteria 5

-- _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ . _ _ . _ _ _ _ . _ . _ -_ ._ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . _ _ _ . _.___.___.___.___J

ATI'ACilMENT A RAI RESPONSES SUBMITTED NOVEMBER 30,1992 (ET-NRC-92-3777)

RAI Number issue 620.27 Editorial, missing a word 620.30 User Behavior / Decision model _

620.31 Modelling process definition _

620.32 Task analysis models _

620.36 Verification and validation _

620.44 Design ITAAC 620.48 SPDS 620.49 Alarm system 720.49 Source term assumptions for MACCS code D

6

- _ _ - _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ - _ ___-_________a

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 100.2 With regard to compliance with the National Environmental Policy Act requirement of severe accident mitigation design alternatives, Section 1.9.5.2 of the SSAR states that the specific AP600 approach will be developed and submitted as part of the desiga certification process, Provide the status of the development and the dsw when this will be submitted.

Response

The AP600 severe accident nutigation design alternatives (SAMDA) evaluation will be subautted by Decemler 15, 1992.

SSAR Revision: NONE W85tingt100S8

NRC REQUEST FOR ADDITIONALINFORMATION Question 100.4 The stalf has not been able to find a list of acronyms and definitions for the terms used for the AP600 design. Nor has it found operational definitions of terms. Provide such a list.

Response

The list of memnyms ca their accompanying definitions are found in Section 1.1.6.5 of the AP600 SSAR titled,*SSAR Acronyms

• starting on page 1.1-2. 'Ihe operational definitions of terms used for the AP600 design are found in Section 16.1.1 titled ' Definitions

• starting on page 16.1-4.

SSAR Revision: NONE

[ W85tingt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION f

Question 210.2 The discussion in Paragraph1 2)(x) oi Section 1.9.3 of the SSAR relative ti testing of relief and safety valves in accordance with item II.D.1 of NUREG-0737 requires additiona! details. The staff concludes that Westinghouse should add the following comustment to this discussion or provide justifration for not doing so:

All of the reactor coolant system relief and safety valves and their assc>aated discharge piping in the AP600 design are similar to those i ems that were tested by EPRI and documented m Reference 2. Any plant specific relief and safety valves and discharge piping that are not similar to thov .ested by EPRI will be tested by the holder of a Combined Operating Liern tn accordance with the ruimes of item II.D.1 of NUREG4737.

Response

The following note will be added to paragraph (2)(x) of Section 1.9.3 of the SSAR:

SSAR Revision:

"The AP600 reactor coolant system design ( , not include power-operated relir .' valves and their associated block vulves. Ilowever, the safety valve and discharge piping used in the AP600 design will be either of similiar design as those items that were tested by EPRI and docutnented in EPRI Report EPRI NP-2770-LD or will be tested in accordance with the guidelines of item [lI.D.1] of NUREG 0737.*

[ W8Stingt10US8

NRC REQUEST FOR ADDITIONAL INFORMATION C.aestion 210.12 Section 3.7.3.9 of the SSAR states that the results of the modal spectrum analysis (mu?tiple input or envelope) are combined with the results from seismic anchor motion (S AM) by the square root sum of the squares netho 1(SRSS).

The ASCE Standard 4-86, ' Seismic Analysis of Safety-Related Nuclear Structures and Comtrentary on Seismic Analysis

• is referenced as the basis for this criterion. He staff has not endorsix! ASCE Standard 4-86, and does not completely .igree with this enterion. For the multiple input or ISM method of analysis, this criterion may lead to unconservative results in some cases. He staff's position, as given in Section 3.9.2.ll.2.g of the SRP, is that the responses due to the inertia effect and seismic anchor motion should be combined by the absolute sum meth(x!.

Provide the technicaljustificatiou for combining the nxxtal spectrum analysis results and the SAM results using the SRSS method.

Response

Technical justification for combining the modal spectrum analysis results and the SAM results using the SRSS method is a comparison with test results as reported in EPRI NP-6153, ' Seismic Analysis of Multiply Supported Piping Systems', Project 964-10, March,1989.

SSAR Revision: NONE WBStingt10USB

)

l NRC REQUEST FOR ADDITIONAL INFORMATION l l

Question 210.13 Section 3.7.3.15 of the SSAR, "An: lysis Procedure for Piping," references the information in Section 3.7.1.3, Table 3.7.1-1, and Figure 3.7.1 13 for cenaia damping values. For the primary coolant loop and other piping systems, ASME Code Case N-411 is referenced in Table 3.7.1 1. Add a note to Table 3.7.1 1 which states that the damping values in Code Case N-411 can be used only as conditioned by RG 1.84. In addition, provide the basis for t'.e 20% damping value which is listed in Table 3.7.1-1 end Figure 3.7.1 13 of the SSAR for 50% to fully loaded cable trays and related supports.

Response

The requirementa of RG 1.84 are satisfied, it is not necessary to add a note to Table 3.7.1-1 since conformance to Regulatory Guides is addressed in SSAR Section 1.9. He damping values used for cable trays and related supports are based on test results. The test results are presented in a report by ANCO Engineers, Inc., " Cable Tray and Conduit Raceway Seismic Test Program, Release 4,* Report #1053-21.1-4, dated December 15, 1978.

The followmg changes will be made to the SSAR:

SSAR Revision:

(ne following sentence will be added to the end of the 2nd paragraph of SSAR Section 3.7.1.3)

"The damping value for cable trays and supports is based on test results (Reference ")."

(The following reference will be added to SSAR Section 3.7)

Reference **, " Cable Tray and Conduit Raceway Seismic Test Program, Release 4,* Report 1053-21.1-4 ANCO Engineers, Inc., December 15, 1978.

3 WBStiR@0USB

NRC REQUEST FOR ADDITIONAL INFORMATION r

Question 210.14 Section 3.7.3.15 of the SSAR states that piping systenu, including coupled equipment, valves and structural frames, can be evaluated with Code Case N-411 damping. Provide the basis for using Code Case N-411 damping values for structural frur.u.

Response

ASME Code Case N411 damping values are not used for structural frames. The SSE damping values for structural frames are presented in subsection 3.7.1.3 Composite modal damping or minimum damping values will be used for analysis models that include piping systems and structural frames.

The SSAR will be revised by replacing subsection 3.7.3.15 with the following:

SSAR Revision:

3.7.3.15 Analysis Procedure for Damping For Damping values used in the seismic analyses of subsystems are presented in Subsection 3.7.1.3.

subsystems other than piping systems that are evaluated with Code Case N-411 damping, and that are composed of different material types, the composite modal damping approach with either the weighted mass or stiffness method is used to determine the composite modal damping value. Alternatively, the mirdmum damping value may be used for these subsystems. Piping systems, including coupled equipment, and valves can be evaluated with Code Case N-411 damping.

210. m 3 Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION i in liii!.

"[

Que".on 220.1 What is the *ASME Design Repoa* and when willit be available for review? Section 3.8.2.1.1 of the SSAR states that *Re information contained in this subsection is based on the design specification and preliminary design and analysis of the vessel. Final detailed analyses willbe documented in the ASME Design Report.* Section 3.8.2.4.1 of the SSAR refers to a

• Preliminary analyses . .* Justify the use of preliminary information in the application for design certification (Section 3.8.2.1.1).

Response

'a ne *ASMS Design Report

• is the report required by paragraph NCA-3550 of the ASME Code. It is prepared by the N Certificate lloider during the procurement and construction phase and is not available as part of the application for design ceitification.

The Code of Federal Regulations, Title 10, Part 52 (10CFR52), paragraph 52.47(2), requires that.. *The informa: ion submitted for a design certification must include performance requiremcats and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the NRC, and procurement specifications and construction and installation specifications by an applicant.*

As described in SSAR Subsection 3.8.2.1.1, the containnvnt vessel is procured as an ASMii metal containment.

' lids procurement includes detail design, fabrication, irutallation, and testing according to il e design specification.

10CFR52 only requires a level of design information sufficient to prepare the piccurement specification. This information is included in the design specification.

The information contained in the SSAR also inO.tes preliminary design and analyses of the vessel. The design described in the SS AR will be provided to the vessel supplier as input to his design. The final design, including any supplier specific details, will comply with the ASME Code and the criteria outlined in the SSAR. Final detailed analyses will be perforum! by the containment vessel supplier and v.ill be documented in the ASME Design Report.

SSAR Revision: NONE 3 W8Stirigt10USB

NRC REQUEST FOR ADDITIONAL.lNFORMATION

..i;ii: liil Question 220.2 The SSAR indicates that the design extemal pressure is 2.5 psig. How has this been considered in the analysis of the containment? What other loadings in Table 3.8.21 of the SSAR are to be combined with the external pressere (Section 3.8.2.1.1)?

Response

The design of the containment vessel includes two hoop stiffeners as shown in SSAR Figure 3.8.2-1, sheets 1 and

4. These stiffeners in combination with the crane girder provide capability in the design for the external differential pressure of 2.5 psid in accordance wid ASME Service Level A limits. The external pressure conditionis combined with dead and live loads during norum operation. The conditions leading to external pressure are discussed in SS AR Subsection 6.2.1.1.2. The critical case is loss of all AC power sources during extreme cold weather. This is evaluated against the external pressure correspondmg to Service Level C of 3.0 psid, The external pre aire loads will be added to SSAR Table 3.8.2-1. The external pressure loading conditions are independent of other extreme events, such as earthquakes.

SSAR Revision:

The SSAR will be revised by replacing Table 3.8.2-1 with the attached table.

Also, the design external pressure will be revised to 2.5 psid to clarify that ii is the differential pressure across the containment boundary.

220.2a Vf westingtmuse

NRC REQUEST FOR ADDITIONALINFORMATION Table 3.8.21 Load Combinations and Service Limits For Containment Vessel Load combination and service limit Load Description Test Design A A C C C D A C x x x x x x x x x Dead D x L x x x x x x x x x 1 Live Wmd W x x x x SSE E, Wg x Tornado Test pressure P x Test temperature T x Operating pressure x Po x x x x x Normal reaction Ro x x x x x Normal thermal To 1 1 1 x Desigt. pressure Pd Pe x

External pressure (2.5 psid)

External pressure (3.0 psid) 1.2 P e x x x x x Accident thermal T. x R, x x x Accident thermal reactions Accideat pipe reactions Yr

• Y- x Jet impingement X

Pipe impact m Notes:

1. Service limit levels are per ASME-NE.

2. Where any load reduces the effects of other loads, that load shall be taken as zero, unless it can be demonstrated that the load is always present or occurs simultaneously with the other loads.

220.. 2 W westingnouse

NRC REQUEST FOR ADDITIONAL INFORMATION N

Ouestion 220.3 There are no shear and tension connectors between the containment vezel and basemat and betwten the cont veasel and internal structures. Vertical and lateral loads on the containment veuel and internal structures are transferred to the basemat below the ves.sel by friction and bearing. Ilow is the potential for relative motion between steel and concrete parts in this region under the various loading combinations considered in the design (Section 3.8.2.1.2)7

Response

The load transfer between the steel and concrete parts within this region under various design load combinations is considered, taking in.o account the steel to concrete interface's ellipsoidal shape and a conservative frictional coefficient of 0.30 between the materials. There is no potential for relative motion between the steel and concrete parts.

SSAR Revisic.i: NONE 22oaa W Westingtiouse

NRC REQUEST FOR ADDITIONAI.INFORMATION iiii! ' 'jit if }{

Question 220.5 Provide a more detailed description of the ultimate capacity evaluation of the cylindrical portion. Describe how strains in the vicinity of local features (such as stiffeners and penetrations) have been incorporated into the analysis.

The area replacement rule raay satisfy strength considerations if sufficient ductility exists, i.e., if locally high straining does not cause premature rupture. How will this be verified for the AP600 containment; i.e., what are the local strain levels at 144 psig (Section 3.8.2.4.2.1)?

Response

ne ultimate capacity of the cylindrical portion is defined as the internal pressure at which gross membrane yielding (based on von Mises criterion) occurs at a critical loca;.an. He critical location is defined as the point which first reaches the yield strength ( See SSAR Section 3.8.2.4.2.1 for more details. ) ne ultimate capacity as defined above is 144 psig. The Service l_evel C maximum pressure is 125 psig. ne ultimate capacity pressure is 15 percent greater than the Service Level C maximum pressure, ne total 4fective ar von Mises strain at 144 psig is the yield strain which is 0.2034 percent. The radial deflection at this pu ure is about 1.6 inches.

local features such as stiffeners and penetrations are designed using details which are permitted in the ASME Code, Section Ill, Subsection NE and experience gained through the fabrication and construction of similar containment vessels. This experience base is enhanced by participation in the '!/8 - Scale Steel Containment Model Pressurized to Failure

• test performed by Sandia National Laboratories and reported in References 220.5-1 and 220.5-2. The results of the 1/8 scale nxxlel test, Reference 220.5-1, indicate that initial fracture was due to a localized detail in which a stiffener discontinuity was located directly adjacent to a shell - reinforcing plate discontinuity. Although this detail initiated failure, it did not affect the overall containment response prior to failure. A localized strain concentration occur.d in the cylindrical shell near the juncture of the stiffeners. The two circumferential stiffeners used in the AP600 are continuous (Figure 3.8.2-1, sheets I and 4 of 4 in the SS AR) and do not intersect penetration nozzles and/or sleeves.

References 220.5-1 and 220.5-2 indicate that the penetrations used in the test vessel were designed in acconlano with the ASME Code rules. The test report shows that the largest measured surface strains of 5.4 percent occurred in the Equipment Hatch penetrution neck close to the reinforcement. In general, strains on the reinforced areas around penetrations were significantly smaller than those in the free fida (2.25 percent) or those in the shell material immediately adjacent to the reinforced areas 0 65 percent ra.uimum). The above listed strain values were recorded at 190 psig. The model rmbrane yield pressure (corresponding to the 144 psig calculated as the ultimate capacity for the AP600) was 155-171 psig and the failure pressure was 195 psig.

The AP600 vessel penetrations were designed to incorporate acceptable details shown in NE-4244. The strain pattern in the AP600 penetrations is expected to be similar to those measured in the test vessel. .

220 m W westinghouse

L NRC REQUEST FOR ADDITIONALINFORMATION I -

REFERENCES 220.5-1: D.B. Clauss, ' Comparison of Analytical Predictions and Experimental Results for a 1:8 - Scale Steel Containment Model Pressurized to Failure," Sandia Natinnal Laboratories. NUREG/CR -

4209, SANDS 5 - 0679, September 1985.

j c

220.5 2: L.N. Koenig, ' Experimental Results for a 1:8 - Scale Steel Model Nuclear Power Plant y Contaimnent Pressurized to Failure," Sandia National bboratories, NUREG/CR 4216, SAND 85

- 0790, December 1986. _

SSAR B; vision: NONE Ti

(

220.5-2 W Westinghouse l

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 220.0 Desenbe the analytical rnodel used in the DOSOR 5 analysis. Justify mesh size (Section 3.8.2.4.2.2).

Response

ne compute cale 'BOSOR-5' was used to determine minimum internal pressure for the nonsynunetric buckling of the top hew. The 'BOSOR-$' aruCysis model with the mesh sizes is shown in attached Figure 220.6-1. Node -

points were concentra'ed in the ' knuckle' region (just above the tangent Ene) where the buckling was expected to occur, he mesh size in this region was less than two times the thickness of the top head. Subsequent analysis verifN that the stresses. strains, and displacements were smooth along the turidional are length. Also, the buckling mode shape was free fromjagged solutions. ne 'BOSOR 5' user's manualindicates that sa absence of jagged solutions is the verification of proper mesh size.

A 100 inch length of the cylindrical shell was included in the anslysis. He analysis sow wi that this Itc;A was adequate to minimize the boundary effects on the buckling of the top head.

SSAR Revision: NONE 1

s W

220.6-1 W-Westirigh0USB

o e NRC REQUEST FOR ADDITIONAL INFORMATION Z

7 = 551.5" ( RAE )

M

[{ .

t* i 025' 73 531 09'

-.i~.

(

2 = 402.19'

! rm Z = 419.05" bi REF ERE NCE SiW ACE ( IN5 s DE ) / [3G Zs 356 "J5" hf ,

( ELL IPSQtDTA56t_ L L ) _.

0I O~~ D " "~

gi 5) R gf e LEGMNT ST ART / END z , 243. -

Q!

l lm OF NODE cue 5H) PClNTS 2 = 193.07"g m Y; IN A SECA4[NT 7m 145.84" h!  !

- w 2 = 100.0" ( T L

• I CYLi p lCAL SHELL Ic g;;

e i.625" li 2 = 0 0"

a. -

INSIT* AADIV5 = 700" f

i /

! DOtt0AAY CONDITION AXlAL AND CIACL8#EPEHTIAL

! DISPLACEuCNTL AND MEAlDIDhAt ROT AT I ON = 0 ( F I XED^),

! AND RADI AL DISPL ACELENT FAEE THI CK NE SS . te 1 625

m1EAlAL- SA337 CL- 2 MATERI AL PAOPERTIES AT Au0 LENT 1FMPE AATUAE MODULUS OF EL Asi t CITY 29 SEE PSt. GOISSON'S AATIO = 0 3.

YIELD STRE CTH

• ED,000 PSI Figure 220.6-1 BOSOR-5 Model of Containment Vessel Head 220.6 2 W- Westin=chouse

NRC REQUEST FOR ADDITIONALINFORMATION I i Question 220.7 Does ' yield' refer to surface stremes or middle surface stresses in Paragraph 2 of Section 3.8.2.4.2.2 of the SSAR'l

Response

Yield refers to the middle surface (membrane) streues. Yield is based on the von Mises criterion.

SSAR Revision: NONE 220.74 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION 1% ly Question 220.13 The buckling factor of safety for the equipment hatch is listed in Secuon 3.8.2.4.2.3 of the SS AR as 1.67, following ASME N-284. What should the factor of safety be fo. Level C7 In Section 3.8.2.4.2 of the SSAR, factors of safety of 2.5,1.5, and 1.67 have been suggested. Justify the selection.

Response

he approach used in Code Case N-284 is based on determining the critical buckling pressure which is equal to the theoretical buckling pressure, reduced by the capacity reduction factor and the plasticity seduction factor. These factors are based on test data. The allowable buckling pressure is the critical buckling pressure divided by the factor of safety. The factor of safety per ASME Code Case N-2&d, Paragraph 1400, is 2.0 for Service Level A/B and 1.67 for Service Level C. His factor of safety of 1.67 was used in the Service Level C evaluation of the equipment hatch.

The factor of safety for Service Level C is 2.5 based on ASME Code,Section III, Subsectiou NE paragraph NE-3222.1. This factor of safety is based on an approach that is not as rigorous as that used in Code Case N-284.

De factor of safety of 1.5 is based on the ALWR Utility Requirements Document and is intended for evaluation of severe accident conditions outside the intended scope of the ASME Code. This criterion is proposed by the industry as an alternate to the proposal in SECY 90-016. The AP600 position is described in Subsection 1.9.5.1

• Containtoent Performance *. He factor of safety of 1.5 is used only for the buckling of the top head subjected to internal pressure. This is justifiable because the probability of significant buckling before tensile yielding of the crown is very low. He following table is a sumrnary of the information shown above:

Table 220.13-1 Factors of Safety for Compressive Simses Service Level ASME NE-3222.1 Code Case N-284 ALWR-URD A 3.0 2.0 -

C 2.5 1.67 -

Severe Arcid.mt - - 1.5 SSAR Revision: NONE

[ WEstingt10VS8

NRC REQUEST FOR ADDITIONAL.lNFORMATION l

Question 220.14 Demonstrate that the equipment hatch seal will not leak at the ultimate capacity. As the containment experiences large strains and displacements, there will tend to be a mismatch of the hatch espe and the cylindrical sleeve. De hatch portion of the seal will tend to displace into a circular shape whereas the cylindrital sleeve portion an elliptical shape, ne two different displaced seal shapes can create a mismatch to remit in seal leakage (Section 3.8.2.4.2.3).

Response

The design of the 22-foot and 16-foot diameter equipment hatches for the AP600 was performed in accordance with the ASME Code, Section !!! sad the experience derived from the test of the 1/8 Scale Model as descrilai in References 210.5-1 and 220.5 7 ~ bis test model included a 1/8 scaled version of a 20-foot diameter equipment batch ( D/t - 60 for model, = ' >r AP600 ) located in the cylindrical portion of a scaled ll2 foot diameter containment vessel. He detail' ac pressure seated equipment hatch were based on those typically used in a full scale unit. He results of the referenced test showed that there was some differential rotation and radial displacement between the hatch sleeve and the cover resulting in slippage at the sealing surface. This combined rotation and slippage was not suf6cient to permit leakage at the rupture pressure. Rupture pressure for the test was 195 psig, which is alumst 5 times the design preasure of 40 psig. He design pressure of the AP600 containment is 45 psig, ne free field strain for the test was reported as 2.25 percent at 190 psig. The test membrane yield pressure was 155 to 171 psig and failv e pressure was 195 psig. He AP600 design assumes the ultimate capacity when the cylindrical portion reaches von Mises yield ( see response to question 220.5 ). For AP600, the ultimate capacity is 144 I sig with the corresponding von Mises or effective strain of 0.2034 percent which is equal to the yield strain.

SSAR Revision: NONE 220 m W westinghouse i

1

NRC REQUEST FOR ADDITIONAL.lNFORMATION Ouestion 220.17 De bearing stress of 33.6 ksf due to the dead load, live load, and safe shutdown earthquake deacribed in Section 3.8.5.5.1 of the SSAR should be included in Table 2.0-1 as the minimum dynamic soil bearing capacity. Modify the table or provide justification for not doing so.

Responso:

ne bearing stress of 33.6 ksf due to the dead load, live load, and safe shutdown earthquake described in 'ection 3.B.5.5.1 of the SSAR is the result of the analysis performed for design of the basemat. The analysis used a conservative approach to bound the range of potential sites. This is described in Subsection 3.8.5.4, wnere it is stated:

" Safe shutdown earthquake loads for the soft rock case, in combination with the pmnerties of soft-to-medium soft soil, are used in the a:alysis since the soft rock case produces higher applied seismic foreca to the structure than the soft to medium soft soil case, llence, the approach is consenative."

The evaluation of the soils and design of soilimprovement, if required, is part of the Combined Licence application and is site-specific. It is conservative to use the reactions calculated from the analyses of the base mat. Ilowever, this is unduly conservative for certain sites. It is sufficient to demonstrate that the bearing reactions for the site-specific soils and SSE are acceptable based on site-specific analyses. As a result, the maximum bearing reaction is not included as the minimum dynamic soil bearing capacity in Table 2.0-1.

The AP600 interfaces for standard design Table 1.0-1, item number 2.13, states that the bearing capacity of foundation materials is a site-specific item to be qualified by the Combined License applicant. As stated in the second paragaph of Section 2.5, the Combined Li- nse applicant must demonstrate that the proposed site "... can support the foundation mat of the AP60C under all specifid site conditions. There is no potential for liquefaction a, ie plant site due to a safe shutdown earthquake." Furthermore, the last paragraph of Section 2.5 provides that

• Bearing loads during seismic conditions for the generic plant are the base reactions from the seismic analyses described in Subsection 3.7.2. The Combined License applicant may either use these loads to demonstrate t. oil bearing acceptability or may perform site-specific seismic analyses to develop bearing loads applicable to his site and seismic conditions.* Finally, the last smnce of Section 3.8.5.5.1 repeats the requirement of "the Combined License applicant will address the interface m Nility of the soil to support the applied foundation loads."

The soil bearing parameter in Table 2.0-1 will be modified to show that the Combined License applicant must address the dynamic capabil"y of the soil to support the applied foundation loads. In addition , the last paragraph of Section 2.5 will be revise.i to reference / accentuate the combined static and dynamic fc.undation stress.

I Westillgt10'JSe i

NRC REQUEST FOR ADDITIONAL INFORMATION SSAR Revision:

De site interface for wils in Table 2.0-1 will be revised as follows:

Soil Bearing Strength Soils must support the AP600 under all specified conditions, he average n.atic bearing reaction of the AP600 is about 8000 pounds / square foot; the maximum static bearing reaction at a corner is 12000 pounds / square foot Shear Wave Velocity Greater t*2an or equal to 1000 ft/sec Liquefaction Pe.ential None The last paragraph of Section 2.5 will be revised to read:

'The average s:stic bearing reaction of the AP600 it bout 8000 pounds / square foot; the maximum bearing reaction at a corner is 12000 pounds / square foot. Bearing loads during seismic conditions for the generic plant are the base reactions from the seismic analyses described in Subsection 3.7.2. He maximum beanng stress due to the dead load, live load, and safe shutdown earthquale is presented in Subsection 3.8.5.5.1 for the worst combir.ation of site and soil conditions. The Combined 1.icense applicaat may either use these loads 9 demonstra9 soil bearing acceptability or may perform site-specific seismic analyses to develop bearing loads applicable tc hMe site and seismic conditions.*

220.17-2 W WeStingh0USS

.r.

NRC REQUEST FOR ADDITIONAL INFORMATION Question 220.18 ne equations with a square root term in Section 3A.3.1.3 of the SSAR sppear incorrect. Cornet or clarify them.

Response

l Le typographical error within the equations in SSAR Section 3 A.3.1.3, sheet P3A-3 will L.e revisal as follows: .

1 SSAR Revision: l I

l 2

Og = { } + - + T 4

se c2

( ) - - + r 4

^

W Westinghouse

NRC REQUEST FOR ADDITloNALINFORMATION M

w IN f

b Cuestion 220.19 Provide the basis for the factors used in defining ellowable streases for the loading ccnditions discussed in Section 3 A.3.1.'t on p. P3 A-3.

Hasponso.

For the nonnal and the severe conditi ns, the towable stress is established, similar to the ANSI Stm lard ANSl/AISC-N600-1984 Wecdon 3A.8, Reference 1, to allow for a factor of safety equal to 1.67. Hence o, a a, / 1.67 = 0.6ay .

For the extnme/=bnormal conditions, the allowable stress is based on the stress limit coefficients in Table Ql.5.7.1 of the same ANSI standard. Hence the alloweble stress is equal to 1.6 times the a, established for the nomial condition or o.96a,.

Tne following will be added to the end of Section 3A.3.1.3:

SSAR Revision:

"For the neanal and the severe conditions, the allowable stress is esiablished, similar to Reference I to allow for a tactor of safety equal to 1.67.11ence a, = a, /1.67 - 0.5ay .

Fur the extreme / abnormal conditiot.s, the allowable stress is based on the stress limit coefficients as Table Ql.5.7.1 of Refescoce 1. llence the allowable strrss is equal to 1.6 times the a, established for the normal condition or 0.96ay .'

l l

l I

1 l

l l

220.19-1 W westinghouSB

NRC REQUEST FOR ADDITIONAL INFORMATION N

Ouestion 220.20 Provide e detailed dewripen and denumstrate the adequacy of the nnharucal connections uxd tojoin a umxtule with reinforcing bars in the concrete (section 3A.5).

Response

ne structural erxxlules are anchored to the concrete using reinforcing steel as shown in Figure 3. A-4. De reinforcing steel is attached to the nnlulea using nuhanical connections wbich consi6 of a $1ceve welded at one ad to the structural nxxtule. The reinforcing steel is anchored to the other end of the sleeve by either an etothertrue process or a threaded connation. De smhanical connections nwet the requirernents of the ACI Cwle, ACI 349-90, (section 3A.8, Reference 2).

ne following paragraph will be added to the end of Section 3A.5:

SSAR Revision:

'De structural nxxtules are anchor-J to the concrete using reinforcing steel as shown in Figure 3. A 4. He reinforcing secci is attached to the unlules using incchanical connections which consist of a sleeve welded at one end to the structural nnlule. De reinforcing steel is anchored to the other end of the sleeve by either an esothermic pmecas or a threaded connection. De mcchanical connections ineet the requirernents of iteference 2.*

I'

'd 220,20-1 W Westinghouse 4

1 NRC REQUEST FOR ADDITIONALINFORMATION

I

p;-

e Question 230.7 Provide the basis of the damping values for cable 'nys, conduits, and their supports prexoted in Table 3.7.1 1 and Figure 3.7.1 13 (section 3.7.1.3).

Response

The damping valuce for cable trays, conduits, and their supports are baud on tests. Please see traponse to Question 210.13.

SSAR Revision: NONE 230.7 1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION i!! "iliti

"~

i Question 230.8 Provide a descripticn and its technical basis for the ** train energy method

• uwx! to model composite damping (Section 3.7.1.3).

Response

The strain energy dependent modal damping valuce are wmputed based on the article by Kross, P,W., *Elenent Anociated Damping by Modal Synthesis,' Proceedings of the Water Reactor Safety Conference, Salt Lake City, Marth 19D, Nstionel Technical Information Servix, U.S. Department of Comrnerce. He modal damping values equals:

{c,}'S,[K,],(d,)

a* . L..,i ( 9, j ' l K, J ( d, )

where:

g, . Ratio of critical da: aping for rnode n nc . Numter of elenents

{ 4, } . Mode n (eigenvector)

[ K, ), . Stiffness matrix of element i g,. Ratio of critical damping associated with element i

( K, ) . Total system stif fness matrix ne following will be added to the end of the third paragraph of Section 3.7.1.3 and Section 3.7.5 references:

SSAR Revision:

• The strain energy dependent modal damping values are computed based on Reference **. The modal damping values equals:

0.M W Westinghouse

NRC DEOUEST FOR ADDITIONAL INFORMATION

{ 4, }' d, [ K,1 { (, )

0* -

E..i (4,)'[K,){4,)

where:

g, . Ratio of critical damping for nale n nc . Number of elemena

( 4, ) . Mode n (eigenvector)

[ K, ), . Stiffnea matrix of element i g, . Ratio of critical damping arveinted with element i

[ K, j . Total system stiffness matrin *

(Reference)

" Krosa, P.W., " Element Associated Damping by Modal Synthesis," Proceedings of the Water Reactor Safety Conference, Salt Lake City, March 1973, National Technical Information Service, U.S. Department of Commerce.

l 230.8 2 W Westinghouse

t'

. , 'k.

NRC REQUEST FOR ADDITIONALINFORMATION 9

!!"i

'ty Question 230.14 liow are 'he truss clenents used in the stick nudet of Figure 3.7.2-4 (Section 3.7.2.1.2)?

\

Hesponse:

De stick nulel in Figure 3.7.2-4 is descrild in SSAR Section 3.7.2.3.1. De s'.ick tmdel considers the eccentricities letween the centroids (the neutral axis for axial and bendmg deformation), the centers of rigidity (the neutral axis for shear and torsional deformation), and the centers of nass of the structures. neae eccentricitica are represented by a combination of tu model elements (one truss and one beam element) in the seismic rmx.lel.

• ne truss element represents only the axial areas of the structural member and is located at the centroid.

His element is developed to resist the verticai seismic input motion.

• ne beam element represent the other structural properties except the axial area of the structural member and is located at the center of ngidity. His beam element is developed to resist the horizontal seismic input motions.

SSAR Revision: NONE 230.1 &1 W westinghouse

e .

NRC REQUEST FOR ADDITIONAL. INFORMATION E

Question 230.16 Deacnbe the method usal to construct a stick avxlel from the axisymmetric shell model of the containment veuel (Section 3.7.2.3.2).

Responsr:

In the stick rrwxlel, the properties were calculated as follows:

1. Members representing the cylindncal portion

- haud on actual circular crou section of the containtnent venel.

2. Member representing the bot *om head

- based on equissient stiffnests calculated from the shell of revolution analyses for st ic 1.0g in vertical and horizontal dirs.. ions.

J. Members representing the top head

- for vertical model, same as 2 ateve,

- for horizontal nxxlel, based on the average of the properties at the successive nodes, using actual cirevlar cross section.

This method used to construct a stick model from the axisymmetric shell nxxlel of the containment veuel is verified by comparison of the nataral frequencies determined from the stick uxxlel and the shell of revolution rmxlel as tabulated in Table 230.16 1.

TABLE 230.161: COMPARISON OF FREQUENCIES MODE NO. VERTICAL MODEL llORIZONTAL MODEL SOR SM SOR SM 1 17.71 18.33 7.39 7.56 2 23.59 30.06 20.88 ,

22.02 SOR: SiiELL OF REVOLUTION MODEL SM: STICK MODEL Prequencica are in liertz.

SSAR Revision: NONE

[ We5tiligh0t!Se

NRC REQUEST FOR ADDITIONAL. lNFORMATION Question 230.19 Ilow is the ' enveloped floor re*Tonse s;wtra' defined? Will they tound the floor response sgctra obtained from the three ac4ign soil profiles? Firure 3.7.2 27 of the $$AR which shows spectrum broadening appem to suggest that only a single floor response spectrum is involved and does not reflect the enveloping procas dexnbod in the lau paragraph of Section 3.7.2.5 of the SSAR.

Response

The ' enveloped floor ressmse spectra' at each unlel nodal point consists of two horirontal and tw vertical llence, response spectra wHch are generated by enveloping the respont.e spectra of the three design soil prr-the enveloped spectra bound the spectie obtained for each of the three design soil profiles. These se eloped floor response spectra are then snmthed and broadened as shown in Figure 3.7.2 27 to form the design floor ressmse spectra used in the design and qualification of subsystems and components.

Figure 3.7.2 27 is intended to show only the procedure used in 6moothing end bioadening of response spectra.

Therefore, it does not reflect /show results of the enveloping proeus. The last sentence of Section 3.7.2.5 will be revi6cd to clarify this intent.

The last sentence of SSAR Section 3.7.2.5 will be revised as shown below.

SSAR Revision:

"Figw*3,M4hw *wwh.h*c-p .x quarum,ilgure 3.7 2 27 shows the srnoothing and bruaderdng procedure tawd to generale the design floor response spectra.'

l 230., S, l w wesnouse

NRC REQUEST FOR ADDITIONALINFORMATION liiH iiii q

Question 230.20 Provide the basis for the third nwthal of Section 3.7.2.6 of the SSAR for combining the reaults from analyzing three compments of earthquake motion. In the line histo y analyses, were all three nwthods used interchangeably to generate a single set of results such as floor rey,nse spectra for all locations of the seisnue Category I structures 7 Response: _

Tic third methal used to consider three uimponents of earthquake notion is basal on the recomnwndation propowd by NUM ARC in Settion '/.7 of NUREO/CR4098, and in Section 3.2.7.1.2 of the ASCE Standard 4 86 (Reference 3),

in the time history analyses, the three methods are NOT used interchangeably to generate a single set of results such as floor resymw spectra for all hwations of the seisnue Category I structurco Only one of the three identified methals is used to generate a single set of seismic responw (2 horirontal & 1 vertical components of responses) for all hwations within thir structure.

SSAR Revision: NONF 230.20 1 W westingtiouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 230.21 The trxxial responwa of the response spectrum analysis of structure *se combined using the square root of the sum of squarea (SRSS) method.1he SRSS method is in agreenent with RO 1.92 if no closely spaced uxxies are preunt. Describe the nethod und for the cases with clowly spaced nxxlea (Section 3.7.2.7),

Response

When closely spxed nx> des are present, these modes are considered using either the grouping method, the ten percent methcx1, or the double sum method shown in Section C of Regulatory Guide 1.92, revision 1.

The following sentence will be inserted after the first sentence of SSAR Section 3.7.2.7 SSAR Revision:

'When closely spaced modes are present, these modes are conddered using either the gn>uping metimd, the ten percent method or the double sum method shown in Section C of Regulatory Guide 1.92, revision 1.*

l [ W85tlagt100SO l

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 231.2 Provide the floor respons.e spatra at the four kotions referencal as the tesis for dennnstrating that the sin seismic conditions are within the AP600 design tusis. His should te documented in the SSAR (Section 2.5).

Response

he floor raponse spectra are currently teing developed. ney will te submitted in April 1993. A reference will be added in SSAR sation 2.5 at that time.

SSAR Revision: NONE 231.2 1 W westinghouse

]

NRC REQUEST FOR ADDITIONAL INFORMATION N

Ouestion 231.4 Table 2.0-1 of the SSAR require 4 minimurn s. oil tearing Strength to be $75 kPa (12 LsO. Provide the basis for accepting a bearing streu of 1610 kPa (33.6 ks0 in Section 3.8.5.5.1 (Section 2.5).

Response

See resionne to Question 220.17.

SSAR Revision: NONE 2314 1 W westinghouse I

NRC REOUEST FOR ADDITIONAL INFORMATION N

Ouestion 231.0 Explain why the item, ' lateral carth pressure loads

• in Table 1.81 of the SSAR is not an item to be addreacd by the combined liceme applicant (Section 2.5).

Response

Lateral earth pre.ssure load is not identified as a site interface item which requires qualification by the Combined 1.icense applicant for the reasons discuued below.

• Lateral carth pressure affects mainly earth retaining structural elements, including retaining walls and structure extenor walls below plant grade.

• ne magnitude of the lateral earth pressure (static and dynamic) is principally a function of the depth of embedment, the location of the water table, the surcharge loads and the level of seismic excitation. ne density and properties of the retsined soil maas is of negligible effect for seismic Category I structurca becam.e of the uniformity among engineering backfills used adjacent to these structures.

• ne earth retaining walls of seismic Category I structures are designed to resist the enveloped lateral earth pressure loads, considering dynamic carth preaw.re h ads due to the safe shutdown earthquake, plus the static carth preasure loads determined assuming maximum soil surcharge load and the worst case external flooding as described in Sution 3.4.

Since the seismic Category I earth retaining structure and below grade exte-ior walls are designed for the worst caw enveloping lateral earth pressure, it is therefore not a site interface requirement.

De following will be added to Section 3.8.4.4.1, second paragraph, after ..

• hydrodynamic, and wind pressure.'

SSAR Revision:

• ne exterior walls of the Seismic Category I structures below the grade are designed to resist the worst case lateral earth pressure loads (static and dynamic), soil surcharge loaas, end loads due *o external flooding as der.cribed in Section 3.4.*

231,01 I

W westinghouse

ulC REQUEST FOR ADDillONAL INFORMATION Ouestion 231.7 State the reavms for not including a discussion in the SSAR of the analysis procedures that would te use1 for evaluating the stability cf d 3 ea, dams, and embankments (Swtion 2.5).

Response

Evaluation of the stability of slopes, dams, and embanknents are site +pecific and are not within the acoge of the AP600 Design Certification. Thew items will te addrened in the Combined Licenw application.

SSAR Rev'aion: NONE 231.7 1 W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION N

Ouestion 231.8 Certain units inny liquefy under vibrettny ground tration. What level and duratica of ground notion is used to au.ea the soil liquefaction potential for the AP600 (Section 2.3)7

Response

The evaluation of wn! liquefaction potentialis site-specific and is outside the wope of AP600 Draign Certification, livaluation of w>il liquefaction will te addreawd in the Cornhined Liceme application using the site-specific safe shutdown earthquake.

SSAR Revision: NONE

~ WO5tingh0VSB

NRC ftEQUEST FOR ADDITIONALINFORMATION Ouestion 231.9 On External Events Analyses, Seismic Margin As;ennat, Appendia 11, a review level casthquake of 0.45 g wp identified for the seismic margin auessment to demountrate sufficient margin over the SSE of 0.30 g. The purpose of the seismic nargins analysis is to test the plant's vulnerability to severe accidents beyond the design basis.

Seismic nargins studies and seismic probabilistic risk asacunats conducted for operating nuclear rower planta have shown the plant IICLPF to le 2 to 3 timna the design value, in view of this, explain why the SSAR chose such a low value for teview level carthquake of 0.45 g, which is only 1.5 times the SSE of 0.3 g7 Responso:

Westinghouse agreca that seismic margin studica as well as neismic probabihstic ri6k asseannats conducted for operating riuclear power plants have shown plant ilCLPP's to b two to three times the design seismic value or higher. Ilowever, the review level earthquake is established to define the screening level used to identify ' weak links' in structures and equipnat. In NUREO 1407 two review or screening level geared to the peak ground acceleration are defined for use when performing a seismic margin evaluation. A 0.3g review level carthquako (RLE) is defined for nmt of the nuclear power plant sites in the Central and Eastem United States (east of the Rocky Mountains). For West Cout sites excluding California coastal sites, a 0.5g review level carthquake is defined, ne AP600 is designed for a Safe Shutdown Earthquake of 0.3g. nis is generally applicable for sites diat are located east of the Rocky Mountains; therefore, a factor relating the SSE and RLE was defined based on factors associated with the NUREO 1407 0.3g bin. Several plants in the 0.3g bin have SSE levels equal to 0.2g. This corresponds to a 1.5 factor, this factor was used to define the 0.45g RLE reported in Appendit if, it is noted that at the tune of the subnuttal of the SSAR, the ALWR Utility Requirementa Document for pauive plants considered a review level car $ quake of 0.43g. Since the SSAR submittal, EPRI has modified the 0.45g level to a 0.5g RLE. %erefore, in future evaluations by Westinghouse the RLE will be defined as 0.5g. Note that mil llCLPF values reported in Appendix 11 are above 0.5g.

Appendix II of the PRA Report willle nulified as follows:

PRA Revision:

(to the fourth paragraph)

A review level carthquake equal to 445 4.50g 2 is established for the seismic margin assessment, and is used to demanstrate sufficient margin over the safe 6hutdown carthquake. him6 ' c4*rt6 uak+*+6w. t mew 4ng-e 4,4-46m-epg44 44+4h.-9 4 c'-d dgn b=! =f+* hut.kewnwathqJe-(OJg),-ni 4e+,n.44. tad-with-th*Mg44n w.644b.WJ by N U F %4mw-bruktwy-Comn4 neon-(NRCHR4ww.w4dio64* t -! fr ;-!r b ga:J far s 0.?g =f:4 *6downweetumke4m4-The seismic margin earthquake that is used is based on Reference 3 median shape spectra, anchored to 0,45t4.50g peak ground acceleration (pga).

[ Westiligt10US0 L

0 NRC FIEOUEST Foil ADDITIONAL INFOllMATION Ii!"

m I.!:

(Fourth senteme of 11.2.4)

• Ihe high confidence of low probabihty of failure valuca are above the review level earthquake (0.4&g0.50g).

(Section 11.3)

The scinnic nurgin study evaluated the AP600 and dermnstrated that the structures, systems, and cornponents required for safe shutdown are enpacted to have high confidenr.c of low probability of failure (ilCLPF) nagnitudc4 greater than the 6 elected review level carthque of 0.4&g0.50g. This provides trurgin above the ufe khutdown earthquake of C 3g, definal am the AP600 site interface.

231.9 2 [ WC5tingh0US0

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 250.1 Section 3.8.2.7 of the SSAR indicates that the innervice inspection of the containment venel will be in ac4crdance mth Subwtion IWE of the 1989 edition of the ASME Section XI Code, llowever. Subsection IWE has tan reviral recently to incorporate operating experience. Therefore, provide information to indicate that the Section XI requirernents are to be augmente=1 with the requirenrnts of Subsection IWE, as revised.

Response

The 1989 Edition of Subsection IWE of ASME Section Xi has tun revisal via the 1990 and 1991 Addenda to incorporate additional inwrvice inspection with NDE requirements applicable to regions of the containirwnt vessel which are accessible from one side only. Sgaifically, Paragraphs IWE 1232. IWE- 2500 and IWE-2600 were reviral to require augumented exanunation of certain areas of the containment vencl which are accessible from one side only. These augmented requirements consist of volunxtric examinations of such areas which must meet the acceptance requirem-nts of IWE 3512.3.

These revimi requirements of ASME Section XI, subsection IWE can te nrt for the AP600 containment vencl.

As discunal in SSAR Subnection 3.8.2.7, the in-service inyection program for the veuel will te descrited in the Combined License application.

SSAn Revision: NONE

[ WOStiflgh0VS8

NRC REOl'E*sT FOR ADDITIONAL INFORMATION

!!! ji tr .g i

Question 250.2 Discuss Westinghouse's propowd procestures in applying the sevised Subsection IWE of the ASME Section XI b de to identify locations in the containment year.cl with propensity for corrosion (Section 3.8.2).

Response

Figure IWE-2500-2 of ASME XI covering exarnination of arras for moisture barriers describes the major area in the AP600 containment vessel which would have a propensity for corrusion. Specific reference is rnade to Table IWE 2500-1, Examination Category E D, note 2, which states *Exandnations shall include internal and external containment moisture barrier materials at conciete-to rnetal interfaces intended to prevent intrusion of moisture against the pressure retaining metal containment shell 0; liner *.

The examinations of these locations will be described in the in-service inspection psogram as part of the Combiaed License application.

Also refer to the restonse to question 250.1.

SSAR Revision: NONE l

250.2 1 l W Westinghouse l

l l

NRC REQUEST FOR ADDITIONALINFORMATION Question 250.3 Demonstrate that all AShiE Cule Class I components will be dcaigned and be provided with access to enable the perfortnance of AShtE Settien XI inspectiota in the instal!cd conditions as required by 10 CFR 50.55a(g). Because the RCPD components will be designed to the 1989 edition,1989 addend., of the AShtE Code as descrital in Section 5.2.1.1 of the SS AR, demonstrate that adequate design and access provisions will te incorporated to pennit inspection for those components that are to juired to te inspected by the 1989 edition,1989 addenda, of the AShlE Snction XI Cale (Section 5.2.4).

Response

AShiE Code Class I components will te designed to that access will te provided in the installed con.lition for visual, surface, and volumetric examinations specified by the AShtE Code Section XI,1989 Edition,1989 Addenda.

Design provisions, in accordance with Section XI Article IWA 1500 will te incorporated in the Class I component design proce,oes.

ne following will be incorporatd . 4to the SSAR, Subsection 5.2.4.2 as the 6rst paragraph:

SSAR Revision:

AShiE Cale Class i componenta are designed so that access is provided in the installed condition for visual, surface and volunwtric exarr.mauons specified by the ash 1E Code Section XI,1989 Edition,1984 Addenda. Design provisions, in accordance with Section XI, Article IWA-1500, are incorporated in the Class I design processes.

250.34 W.

Westincrhouse

d NRC REQUEST FOR ADDITIONAt. lNFORMATION 4

1 Ouwdon 2fa0.4 DetnorWrcte tbt the preservice inspmtion(PSI) of all ASME Code Class I components wi:1 meet tb 1989 edition, 1989 addenda. of the ASME Section XI Cale as required by 10 CFR 50.55alg). Because die PSI raiuirements have teen cua' usbs!,10 CFR 50.55a(g) does not have provisions for relier requwu for impractical PSI exannestion %uirements. P ovide ir formation to confitm that all PSI requirenwnts will be nwt (Section 5.2.4).

RMporbo:

Prewrvice inspection is a combined license applicant interfect as listed on Table 1.8.1.

SSAR Subuxtion 5.2.4 Par Frsph I will te modified as follows:

SSAR Revisior,:

Preservice and inservice inspection and testing of ASME Code Class I prer.aute-setnining components (including venels, piping, pumps, valves, botting, and supporta) within the reactor coolant preasure boundary are perfornvxt in accordance with Section XI of the ASME Code including addenda according to 10 CFR 50.55a(g). This includes all ASME Code Section XI Mandatory Appendices.

o 2soa.,

w weSunciouSe

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 250.5 ASM E Section XI indicates that the PSI 6hould be conJucted with equipment and techniques equivalent to tho6e that are espected to le uwd for subuquent intervice inspec'.jon (ISI). He PSI provides the baseline information for reference in subsequent 151. For example, if the 151 ef piping weld i' expected to be lwrfornal with ultraumic techniques, the PSI 6hould ab.o be bawd on ultraumic techniques. Provide information to cemfirm that this raguirenwnt will be satisfied for all ASME Code Clau I cornpanents (Section $.2.4).

Ac:acn*,c:

Preunice inste; doc is o enmbined licene applicard interface as listed on Table 1.8.1.

The prewnice inspection (PSI) will be conducted with equiprnent and techniques equivalent to those that are expected to be used for subwquent inservite inspection (ISI). This is an ASME Code Section XI requirement. The PSI and ISI are to te performed in accordance with the ASME Code,Section XI. nis is stated in the resp (mse to question 230.4.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 250.0 Provide information to confirm that Article IWA 1500, ' Accessibility,* of Section XI of the ASME Code will be satisfied for all ASME Code Class I conyonents (Section 5.2.4).

Response

Design provisions, in accordance with Section XI Article IWA 1500, willic incorporated in the Class I component design proceses. The response to question 250.3 contains this statement.

SSAR Revision: NONE l

250.M W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION n,,,

,,g Ni Ouestion 250.7 The ASME has published Appendix Vil, ' Qualification of Nondestructive Examination Personnel fo/ Ultrasonic Examination,' and Appendix Vill,

• Performance Demonstrstion for Ultraxonic Examination Systems,' in Section XI (Division 1) of the ASME Code. He NRC has published (m the Federal Register)its intent to reference in 10 CFR 50.55s(b) the ASME Section XI edition that Mcludes the published Appendix Vi! In addition, the NRC staff has established a technical contact to coordinate the implementation of Appendix Vill. Herefore, indicate that Section XI requirements are to be augmented with the requirements in Appendices Vil and Vllt for all ASME Code Clau 1 components (Section 5.2.4).

Response

Qualification of nondestructive examination perr.onnel and the performance demonstration of examination systems in accordance with ASME Section XI are a combined license applicant interface as listed on SSAR Table 1.8.1.

All mandatory appendices are required by the ASME Code,Section XI. Appendix Vil is mandatory and Appendix Vill i.a intended to be mandatory. The response to question 250.4 is intended to include these appendices.

SSAR Revision: NONE

[ WCStingh0USB

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 250.8 ASME Code Class 1, 2, and 3 carbon and low-alloy steel piping items that are susceptible to wall thinning as a result of the single-phuc (water) erosion / corrosion phenomenon will be subject to examination in accordance with Subsection IWil of ASME Section XI. Therefore, indicate that Section XI requirements are to be augmented with the requirements of Subsection IWil for all ASME Code Clus I components (Section 5.2.4).

Response

ne AP600 Class I piping is fabricated of stainless steel and not suxcptible to wall thinning by erosion / corrosion. The primary components in contact with reactor coolant that are fabricoted of low alloy steel are clad with corrosion resistant material and are not suxeptible to wall thinning by erosion / corrosion. AlthouEh the secondary side of the steam generator is constructed to the requirements for Class I components it is considered to be a Class 2 component for the purposes of inservice inspection programs.

Subsection IWil of the ASME Code,Section XI has not been approved at this time. A commitnwnt in the SSAR to use a portion of the ASME Code not yet approved not authorized for use by the NRC is inappropriate.

Additionally, the preservice and inservice inspection plans are not part of the Design Certification but are to tw prepared by the combined license holder in compliance with the requirements of 10 CFR 50.55a.

SSAR Revision: NONE

[ WOStiflgh0US0

NRC REQUEST FOR ADDITIONAL. lNFORMATION N

Question 250.22 Demonstrate that all ASME Code Clau 2 and 3 congonents will be designed and te provided with accca to enable the performance of ASME Section XI inspections in the installed conditions as required by I? CFR 50.55a(g).

Further, confirm that Class 2 and 3 components will be designed to the 1989 edition,1989 addenda, of the ASME Code. Verify that the applicable ingations are those in the 1989 edition,1989 addenda, of the ASME Section XI Code (Section 6.6).

Response

ASME Code Class 2 and 3 components will be designed so that access will be provided in the installed c.onditioa for visual, surface and volumetric examinstions specified by the ASME Code,Section XI,1989 Edition,1989 Addends. Design provisions, in accordance with Section XI, IWA 1500, will te fonnally implemented in the Clam 2 and 3 component design processes.

He following will be incorporated into the SSAR, Subsection 6.6.2 as the first paragraph.

SSAR Revision:

ASME Code Class 2 and 3 components are designed so that access is provided in the installed sandition fc.r visual, surface and volumetric examinations specified by the ASME Code, Secuon XI,1989 Edition,1989 Addenda. Design provisions, in accordance with Sect.on XI, IWA-1500, are formally implernented in the Class 2 and 3 component design processes.

250.22 1 W westingtiouse

I 8 NRC tlEQUEST FOR ADDITIONALINFORMATION I

Ouestion 250.23 Demonstrate that the preservice inspection (PSI) of all ASME Code Class 2 and 3 ceniponents will nwt dts 1989 edition,1989 addenda, of the ASME Section XI Oxle as required by 10 CFR 50.55r(g). Becam.e the PSI requitetnents have leen established,10 CFR 50.55a(s) dca not have provisions for relief requests for impiactical

, pSt esamination requirenwnts. Provide information to conGrm that all PS! requirements will be met (Section 6.6).

Response

Preaervice and inwrvice inspections of Quality Group D and C pressure retaining components ( ASME Code Section all Class 2 and 3 components ) such as vessels, piping, pumps, valves, botting, and supports idemined in SSAR Subt.ection 3.2.2 will be performed in accordance with the ASME Code,Section XI, as required by 10 CFR 50.55s(g). This includes the ASMU Code Section XI Mandatary Appendices.

SS AR Subsection 6.6.1, first paragraph, will be changed as follows:

SSAR Revision:

Prese:vice and inservice inspections of Quahty Group D and C pressure retaining components ( ASME Code Section til Class 2 and 3 components ) such as vessels, piping, pumps, valvos, bolting, and supports as identified in Subsection 3.2.2 are perfonned in accordance with the ASME Code,Section XI, as required by 10 CFR 50.55a(g).

This meludes the ASME Ode Section XI Mandstory Appendices.

WB5tingh0058

r NRC REQUEST FOR ADDITIONALINFORMATION Question 250.24 ASME Section XI indicates that the PSI abould be conducted with equipment and techniques equivalent to those that are expected to be used for subsequent inservice inxpection (151). He PSI provides the baseline information for reference in subsequent 151. For example, if the 151 of piping weld is capected to be perfornwd with ultrasonic techniques, the PSI should also le nased on ultrasonic techniques. Provide information to confirm that this requirens.nt will be natisfied for all ASME Code Clus 2 and 3 components (Section 6.6).

Response

i Preservice inspection is a combined license applicant interface as listed on SSAR Table 1.8.1.

The preservice irwpection of Class 2 and 3 components will be conducted with equipment and tecimiques equivalent to those that are expected to be used for subsequt inservice inspecthm ( 151 ). This is an ASME Code Section XI requirement. The PSI and 151 are to be pcrformed in accordance with ASME Code Section XI His is stated in the response to question 250.23.

SSAR Revision: NONE 2so.2m w westinghouse

1 i

i NRC REQUEST FOR ADDITIONALINFORMATION l

i Question 250.25 Provide infornation to confirm that Article IWA 1500, 'Acceuibility,* of Section XI of the ASME Code tvill be satisfirx! for all ASME Code Clars 2 and 3 cornponents (Section 6.6).

Response

Design provisions, in accordance with Section XI Article IWA-1500, will be incorpensted in the Class 2 and 3 component d-sign processes. The respone to question 250.22 contains this 6tatement.

SSAR Revision: NONE M WB5tingh0USB

NRC REQUEST FOR ADDITIONAL INFORMATION I

Question 250.26 The ASME has published Appendix VII, ' Qualification of Nondcatructive Exatnination Perwnnel for Ultrasonic Examination,'snd Appendix Vill, 'Pe4 rmance Denonstration for Ultrasonic Examination Systems," in ASME Section XI (Division 1). The NRC has pt.olished in the Federal Register its intent to refuence in la CFR 50.55a(b) the ASME Section XI edition that includes the published Appendix Vil, in addition, the NRC staff has established a technical contact to coordinate the implementation of Appendix Vill. Therefore, provide information to indicate that Section XI requirements are to be augmented with the requirements in Appendices Vil and Vlll for all ASME Code Class 2 and 3 omponents (Section 6.6). _

Resporise:

Qualification of nondestructive examination personnel wd tin performance demonstration of examination systems in accordance with ASME Section XI are a combined license applicant tnterface as listed on SSAR Table 1.8.1. "

All mandatory appendices are required by ASME Code Section XI. Appendix Vilis mandatory and Appendix Vill is intended to be mandatory. The response to question 250.23 is intended to include these appendices.

SSAR Revision: NONE 2so.2s.,

w wesunsouSe

NRC REQUEST FOR ADDITIONAL. lNFORMATION k

Ouestion 250.27 ASME Cc e Class 1,2, and 3 carbon and low-alloy steel piping items that are susceptible to wall thinning as a result of the single-phase (water) crosion/ corrosion phenomenon will be subject to examination in accordance with Subsection IWH of ASME Section XI. Therefore, provide information to indicate that Section XI requirements are to be augment d with the requiremv..ts of Subsection IWil for all ASME Code Class 2 and 3 components (Section 6 6).

Response

ASME Section XI examination requirernerits are a combinu! licens . ,... cant interface as listed on SSAR Table 1.B.1.

Subsection IWil of the ASME Code,Section XI has not been approved at this time. A commitment in the SSAR to use a portion of the ASME Code not yet approved nor authorind for use by the NRC is inappropriate.

Additionally, the preservice and int.ervice inspectica plans are not part of the Design Certification but are to be prepared by the combined license bolder in compliance with the requirements of 10 CFR 50.55a.

SSAR Revision: NONE

[ W8Stingh00Se

NRC REQUEST FOR ADDITIONAL INFORMATION

!Hi! i! i W  !.]

Ouestion 250.23 If drilled holes will be present in the rotor, discuss the preservice inspection requirements for them (Section 10.2.3).

Response

During the manufacture of rotors, a variety of holes are drilled into the rotor todies. Included are: coupling tol*

holes, jackscrew holes, balance plug holes and equilibrium holes, he penetrations are inspected for surface finish, sin, and Apth. NDE inspection is performed on the holes after final machining of the rotor and prict to the drilling and tapping operations.

'Ibe AP600 turbine factory inspection prograin is outlined in the Table 250.28-1. Note that during the factory inspections both the IIP and LP turbines are subjected to non-destructive examination of the surfaces. Thise inspections are identif ed as Customer Witness Points and the test results recorded in the permanent Quality Records.

SSAR Revision: NONE l

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25o.28 1 W westingtiouse l

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

s NRC REQUEST FOR ADDITIONALINFORMATION TABLE 250.28-1 ,

FACTORY INSPECTION AND TEST PROGRAM '

HP TURBINE HP ROTOR FOROINGS '

At the supplier, the following are approved and/or witnessed:

ladle enalysis. (R)

Ultrasonic examination after heat treatment for properties and skim turmng. (R)

Rotor is heat treated. (R) '

Mechanical property teat. (R)

Thermal stability test. (R)

Dimensional checks. -

In the Factory:

Finish machined dimensional check Concentricity check Nondestructive examination of surfaces. (W/R)

Side-entry milling or "T" slot machining inspection-Blade assembly check.

Shroud assembly check.=

Check hardness of nvet heads (hot riveted).

Check shroud tightness on blades, river condition, blade locking, and seal strip caulking _

Check rotor balance at design and overspeed.* (W/R)

Check shrouds and rotor for concentricity after overspeed Check shroud gaps Verify proper release from manufacturing / assembly area, check preservation and preparation for shipment and release for loading on carriers. (W/R) .

HP Outer and Inner cylinders At the supplier, the following are approved and/or witnessed:

- Ladle analysis. (R) i Magnetic particle examination. (R) -

(R) Quality Records (W)- Customer Witness Point

• - Test equipment adjustment sometimes required -

250,28-2 W Westinghouse

- . , . . . , =. .. , ::

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NRC REQUEST FOR ADDITIONAL INFORMATION lHi! lii '

if TABLE 250.28-1 FACTORY INSPECTION AND TEST PROGRAM IIP TURBINE LP ROTOR FORGINGS At the supplier, the following are approved and/or witnessed:

Ladle analysis. (R)

Ultrasonic examination after heat treatment for properties and skim turning. (R)

Rotor is heat treated. (R)

Mechanical property test. (R)

Thermal tability test. (R)

Dimensior.al checks.

In the Factory:

Finish machined dimensional check Concentricity check Nondestructive examination of surfaces. (W/R)

Side-entry milling inspection Blade assembly check Shroud assembly check.

Check shroud tightness on blades, rivet condition, blade locking, and seal strip caulking.

Gage blade steam passages.

Check shruud machined dimensions Nondestructive examination of lashing lug welds.

Rotating blade group stationary frequency test.

Check shroud, seal machined dimensions.

Check rotor balance at design and overspeed.* (W/R)

Check shrouds and rotor for concentricity after overspeed.

Non-destructive examination of erosion shields.

Verify proper release from manufacturing / assembly area, check preservation and preparation for shipment and release for loading on carriers. (W/R)

(R) - Quality Records (W) - Customer Witness Point

• - Test equipment adjustment sometimes required 2so.28-3 w westineouse

NRC REQUEST FOR ADDITIONAL INFORMATION I

Question 250.29 Provide information to confirm that the inservice inspection program discussed in Section 10.2.3.6 of the SSAR will d

ensure that the failuse and missile generation probability will be less than 10 per year. (See Q251.1)

Response

He inspection requirements listed in SSAR paragraph 10.2.3.6 are well within the instation guidelines noted in Reference 1 of SSAR Subsection 3.5.5. He following conclusion is presented on page 44 of the reference:

• As with previous designs, the potential for SCC has the greatest influence on rotor integrity llowever, in fully integral designs the probability of failure by this mechanis:n has been reduced drastically and the analysis shows that 30 years of running time, or more, may elapse before inspection without exceeding the NRC safety criteria.'

SSAR Revision: NONE WBStingt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION Question 251.1 he staffs position regarding turbine maintenance and inspection is that the turbine maintenance and inspection program be implemented to ensure that the failure and missile generrtion probability is less than 10d per year for a favorably oriented turbine system (see letter from C. E. Rossi (NRC) to J. A. Martin (Westinghouse) dated February 2,1997]. Describe how this position will be met (Section 3.5.1.3).

Response

The turbine is favorably oriented and the unit maintenance and inspection program proposed for AP600 is conservative.

Please refer to our comment to RAI Item 250.29.

SSAR Revision: NONE W Westinghouse

l l

NRC REQUEST FOR ADDITIONAL. INFORMATION I

Question 251.24 Section 10.2.3.2 of the SSAR indicates that flaws may be acceptable in the rotor if the flaws can be shown not to grow to critical sius. A lisw growth evaluation to demonstrate structural integrity in lieu of flaw removal is not consistent with the ASME Section 111 Code which does not permit a flaw evaluation. Discuss how the acceptance criteria in Section 111 and Section V of the ASME Code are met.

Response

ne design, manufacture and inspection of the steam turbine does not fall under thejurisdiction of the ASME code.

Identified sinel e : laws in die rotor are evaluated for the potential to grow to critical within the design life. WSTG-4-P, Sect. 4 (Reference 1. SSAR Section 3.5.5) discusses the methodology for estimating the growth of internal flaws. For evaluation, upper bound flaw growth rate parameters are used in the Paris equation. He critical flaw sia is estimated from the stress and fracture toughness. An initial sia corresponding to the duty cycle is calculated frorn Eq. 4.3. The initial sin so found must correspond to an g a that is 20 times the UT reported area for that indication to be acceptable. Combinations of indications are also evaluated by complex grouping procedures.

SSAR Revisc.,n: NONE 251.2 & 1 W Westinghouse

)

l

NRC REQUEST FOR ADDITIONAL INTURMMION Ouestion 251.25 Demonstrate that the fracture appearance transition temperature (50% FATT) as obtained from Charpy tests performed in accordance with ASTM A370 will be no higher than 18'C (O'F) for low-pressure turbine rotors (Section 10.2.3).

Response

Westinghouse material procurement specifications require FATT development via Charpy tests according to ASTM A 370 and that the FATT not be higher than -18 'C (0 'F). Figure 8 of the Reference I shows test data from producuan rotor forgings where the FATT tested less than 0 'F.

REFERENCE:

1. Argo, H. C., Novak, R. L., Westinghouse Electric Corporation, Power Generation, Orlando Florida, ' Material Property data of large Integral Low Pressure Rotor Forgings Designed for Improved Reliability and Efficiency,"

10th International Forging Conference, September 23-25,1985, University of Sheffield, England.

SSAR Revision: NONE 2s1.2s-1 W westingtiouse

NRC REQUEST FOR ADDITIONALINFORMATION Question 251.26 Provide information to show that the Charpy V-notch eneigy at the minimum operating temperature oflow-pressure rotors in the tangential direction will be at least 82 J (60 ft-lb) (Section 10.2.3).

Response

The Charpy energy of the material for the LP rotor must be 70 ft-lbs minimum at the bore. This is also the upper shelf energy Samples of the material are taken from the bore of each forging to validate specifi.2 tion conformance.

SSAR Revision: NONE W85tiligt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION Question 251.27 Provide information in Section 10.2.3.2 of the SSAR to demorwtrate that the ratio of the fracture toughness "Kc' of the rotor material to the maximum tangential stress at speeds from normal to design overspeed will te at least 3.2 Vem (2 Vin), at minimum operating temperature.

Response

The fractme toughness of the most highly stressed e.rea of the rotor, estimated from the Charpy energy by means of the Rolfe-Novak equation and applying extra conservatism (RAI 251.29), is approximately 149 KSt. The maximum stress in this area at design speed is 47.8 KSI. 'Ibe ratio of Ke to Stress is 2.92 Vin and at 20 percent overspeed, the ratio is 2.04 Vin.

The third sentence in the Section 10.2.3.2 of the SSAR will k revised as follows:

SSAR Revision:

The ratio of material fracture toughness, K,e (as derived from material tests on each rotor) to the maximum tangential stress for rotors at speeds from normal to design overspeed killlie at least 2 Vin'at mirLimunf operatiiig temperature.

2s1.27-1 W westinghouse l

NRC REQUEST FOR ADDITIONALINFORMATION ,

l Question 251.28 Provide information to show that sufficient warmup time will be specified in the turbine operating instructions to ensure that toughness will be adequate to prevent brittle fracture during <tanup (Section 10.2.3).

Response

The maximum toughness of the mkierial is developed at room temperature. No rotor warming is necessary.

SSAR Revisiott NONE

[ W85tingh0USS

NRC REQUEST FOR ADDITIONAL INFORMATION h "il!.

!l Question 251.30 Section 10.2.3.4 of the SSAR indicates that the low-pressure turbine element hu a central bore while the high-pressure turbine element does not. De staff considers a central bore desirable to remove impurity inclusions from integral rotors. Provide technicaljustifications for not boring the high-pressure rotor.

Response

llistorically, IIP and LP turbine rotors, for all applications, have been bored to remove impurity inclusions, it has been demonstrated that in the sizes of rote used for llP turbines, boring is no longer necessary from the standpoint of safety. This is based on several reasons:

• Improved rtitor quality.

• Improved inspection and quality condition characterization by ultrasonic techniques.

• Improved und.erstanding of the linear clastic fracture mechanics pbcnomena.

• The low level of stress in the half speed (1800 rpm) turbines.

The use of unbored rotors for llP turbines, of current design sizes is now a general industry practice.

SSAR Revision: NONE 2s1.30 1 W westingticuse

NRC REQUEST FOR ADDITIONALINFORMATION i

Question 251.31 Confirm that each fmished rotor will be subjected to 100% volumetric (ultrasonic), surface, and visual examinations using procedures and acceptance criteria equivalent to those specified for Class I components in Sections III and V of the ASME Code (Section 10.2.3).

Response

ASME Code Section III requirements do not apply to turbine rotors. The rotor forging will receive a 100 percent volumetric inspection in the drum state, prior to acceptance from the supplier. This inspection performed by Westinghouse inspectors is in accordance with Westinghouse requirements.

The configuration of the rotor in the finish machined condition makes a 100 percent volumetric inspection extremely difficult and unreliable. Periodic inspection of the near bore region may be performed if such need were demonstrated. Westinghouse Report WSTG-4-P (Reference 1, SSAR Section 3.5.5) to the NRC concluded that such inspections are not necessary to satisfy the required probability levels in the design life.

SSAR Revision: NONE W85tiflgt100SB

NRC REQUEST FOR ADDITIONAL INFORMATION i  !

Question 252.20 Section 3.8.2.6 of the SSAR indicates that the containment veuel materials will be impact tested according to Article NE-2000 of the ASME Code. However, Section 6.2.7 of the SRP recommends that the fracture toughness of the reactor containment pressure boundary materials should meet the fracture toughness requirements in Subsection NC of the ASME Section 111 Code. Provide techmcaljustifications for this deviation.

Response

The jurisdictional boundaries for the AP600 containment vessel and associated piping have been established usmg Figure NE-Il20-1 of ASME !!!, Subsection NE. He fracture toghness requirements for those items contained within the Class MC boundaries are in accordance with the requirements of NE-2300. The fracture toughness requirements for those items contained within the Clasa 2 piping system boundaries are in accordance with NC-2300.

A comparison of these two sets of requirements shows that they are essentially the same, therefore, the containment vessel material meets the fracture toughness requirements of Class 2 piping.

SSAR Revision: NONE I

252.20 1

[ Westingh00Se l

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. , 1 NRC REQUEST FOR ADDITIONAL INFORMATION rrn h

(

Ouestion 252.36 Section 4.5.1.1 of the SSAR indicates that materials in the control rod drive system are selected based on certain number of plant transients. For example, the SSAR assumes 320 reactor trips, liowever, the Standard Technical Specifications list 500 reactor trips for a plant with a 40-year design life. Demonstrate that the assumed plant transients in the SSAR are applicable to the projected 60-year plant design life (Section 4.5.1).

Response

ne AP600 Reactor Coolant System Design Transients are described in SSAR Subsection 3.9.1.1 and are included in Table 3.9-1. The number of occurrences of each transient were derived from the Westinghouse NSSS Design Transients for current plants. For AP600, the number of occurrences of each transient were then adjusted based on the following:

- AP600 60 year design life

- AP600 design features

- IIistorical Westinghouse plant performance

- Projected AP600 plant performance (reactor trips, availability)

SSAR Table 3.9-1 specifies 300 reactor trips. This is based on 3 reactor trips per year from low power, and 2 reactor trips per year from full power. His is a conservative assumption considering the design features incorporated in the AP600 to reduce reactor trips including:

- Larger pressurizer

- Full (100 %) load rejection capability

- Automatic feedwater flow control

- Two out of four protection logic The current estimate of reactor trips for the AP600 is approximately 1 per year. Herefore, the assumed total number of occurences of reactor trip (5 per year) is conservative.

SSAR Revision:

Subsection 4.5.1.1 of the SSAR will be changed to indicate that the analysis of the CRDM's will assume 300 reactor trips.

1 252.au W westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION l

Question 252.40 Section 4.5.1.3 of the SSAR indicates that Inconel 750 materials to be used in the control rod drive system will te ordered to specifications other than those in ASME So: tion Ill. Provide technicaljustifications that the alternative specifications nwet the requirernents in ASME Section III.

Response

"Ihe nickel-chromium-iron alloy referred to in the subject section is not part of a pressure boundary nor is it welded to a pressure boundary. Therefore ASME material specification requirements are not required. The material specifications used are sufficient to prevent the introduction of contaminants into the reactor coolant system. The capability of springs fabricated of this material to meet functional requirements is based on many years of satisfactory operation in this application. This operational experience represents the technical justification for the use of this alloy. Please note that the operating temperature of the springs is typically below that of the bulk of the reactor coolent, therefore, the potential for primary water stress corrosion cracking is minimized.

SSAR Revision: NONE 252.40-1 t

l [ WB5tingt101]Se

NRC REQUEST FOR ADDITIONAL.INFORMATION Ouestion 252.42 Because Section 4.5.2 of the SSAR discusses both reactor internal n.nj n core support materials, consider revising the title of Section 4.5.2 of the SSAR accordingly.

Response

Although core supports are a subset of reactor internals, the title of this section will be changed to " Reactor Internal and Core Support Materials

• to be consistent with the Standard Review Plan.

SSAR Revision:

4.5.2 Reactor Internale and Core Support Materials 4.5.2.1 Materials Specifications The major material for the reactor internals is..

2s2A2-1 W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 252.57

_ E ,

Section 5.2.3.3.1 of the SSAR indicates that the fracture toughness properties' of the'RCPB may meet the requirements of the ASME Code,Section III, Subsection NC. However, Subsection NC is for ASME Cale Class 2 components, and is not appuuble to the RCPB. Clarify your intent relating to the application of Subsection NC -

in the RCPB.

Response

he requirements of ASME Code,Section III, Subsection NC apply to those portionsof the' reactor coolant pressure boundary that are not required to meet the requirements for Class I components as determined by the criteria in 10 CFR 50.55a(c)(2). Those portions of the reactor coolant pressure boundary that are not Class 1 are Quality Group B in confonnance with Regulatory Guide 1.26. The first paragraph of SSAR Sunsection 5.2.3.3.1 will be revised to clarify this issue as follows:

SSAR Revision:

'The fracture toughness properties of the reactor coolknt pressure boundary components meet the requirements of the ASME Code,Section III, h=;;;E Subarticle NS-2300 =' MC. nose portions of the reactor ' coolant pressure boundary that meet the requirements of ASME Code,Section III, Clasa 2 per the criteria of 10 CFR 50.55a, meet the fracture toughness requirements of the ASME Code,Section III, Subarticle NC-2300."

1 9

i 252.s7-1 W . westinghouse l -

I L _,

NRC REQUEST FOR ADDITIONAL.lNFORMATION 3?)li a:

- ~

Question 252.91 Section 5.3.5 of the SS AR indicates tiat the reactor vessel materials surveillance program will be in accordance with ASTM E185-il3. Ilowever, the applicable version of ASTM E185 that is referenced in Appendix 11 to 10 CFR Part 50 is ASTM E185-82. Demonstrate how Appendix 11 to 10 CFR Part 50 is met.

Response

Reference Number 1 ' ASTM E-185 83* is in error. The correct reference number 1 is:

SSAR Revidon:

• 1. ASTM E-185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactar Vessels.'

M 252.91-1 W Westingt100Se l

NRC REQUEST FOR ADDITIONALINFORMATION f

Ouestion 252.97 Discun design provisions for the installation of trplcement surveillance capsules (Section 5.3).

Response

As stated in the SSAR test (Subsection 5.3.2.6 Material Surveillance) archive material sufficient for two additional capsules will be retained.

• The surveillance capsules are designed to maintain the test specimens in an inert environment within a corrosion-resistant capsule to prevent deterioration of the surface of the specimens during radiation exposum. He design of the surveillance capsules is such that the temperature history of the specimens duplicates, as closely as possible, the temperatur* cxperienced by the reactor vessel. The surveillance capsules are designed to be sufficiently rigid to prevent mechanical damage to the specimens and monitors during installation and irradiation.

The design of the surveillance capsules and capsule attachments are based on the current proven Westinghouse design and permits insertion of replacement capsules into the reactor vessel if required at a later time in the lifetime of the vessel.

He design of the capsule holders and the means of attachment 1) preclude structural material degradation by the attachment welds,2) avoid interference with inservice inspection required by ASME Code Section XI, and 3) ensure the integrity of the capsule holder during the service life of the reactor vessel."

GSAR Revision: NONE

[ WOStiflgt100S8

NRC REQUEST FOR ADDITIONAL INFORMATION N

Question 252.98 Section 5.3.4.6 of the SSAR indicates that there is a Table 5.3-7 in the SSAR. He staff cannot fmd this table.

Correct or clarify this reference.

Response

The reference to SSAR Table 5.3-7 will be deleted. SS AR Table 5.3-7 will be deletal and the fifth paragraph of Subsection 5.3.4.6 revised to reflect this response as follows:

SSAR Revision:

ne reactor vessel beltline materials are specified in Subsection 5.3.2. He fluence of 2 x 10 l9 n/cm whichis the 2

design basis fluence at the vessel inner radius, at 54 EFPY, at the pd location, was used for calculating the i

RTyrs value. RTjrFS s RTNDT, the reference nil ductility transition temperature as calculated by the method chosen by the NRC staff as presented in paragraph (b)(2) of 10 CFR 50.61, and the pressurized thermal shock rule.

The pressurized thermal shock rule states that this inethod of calculating RTPTS should be used in reportin; values used to compare pressurized thermal shock to the above screening criterion set in the pressurized thermal shock rule.

He screening criteria will not be exceeded using the method of calculation prescribed by the pressurized thermal shock rule for the vessel design objective. The material properties, initial RTNDT, and end-of-life RTFTS V*l" are in Tables 5.3-1 and 5.3-3. He n= :d6 E'reific' i- T 5!: 5.2  ::c i== materials that are exposed to high fluence levels at the beltline region of the reactor vessel emi-arc, i=:fn::, 6: subject to 4the pressurized thermal shock rule. These materialsr-th= %::, are a subset of the reactor vessel materials identified in Subsection 5.3.2.

2s2ssa W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION I f Question 252.118 Section 6.1.1.1 of the SSAR discusses ' principal" materials for the enginected safety features (ESF). Provide information on all materials in the ESF.

Response

Secuon 6.1.1.1 of de SSAR discusses the ' principal

• pressure-retaining materials in engineered safety features system components. In Subsection 6.1.1.1 of the SSAR, the word ' principal" will be removed and replaced with

• the*, since only those materials which are pressure retaining are discussed. Also the title to Table 6.1-1 will be revised accardingly.

SSAR Revision:

6.1.1.1 Specifications for Pe+pel the Pre.sure-Retaining Materials Peipal The pressure-retaining materials in engineered safety features system components comply with the corresponding material specification permitted by the ASME Code, Section 111, Division 1. The material specifications for pressure-retaining materials in each component of an engineered safety features system meet the requirements of Article NC-2000 of the ASME Code,Section III, Class 2, for Quality Group B; Article ND-2000 of the ASME Code,Section III, Class 3, for Quality Group C components; and Article NE-2000 of the ASME Code,Section III for containment pressure boundary components. Materials produced under ASTM designation are acceptable as complying with the corresponding ASME specification, provided the ASME specification is designated as being identical with the ASTM specification for the grade, class, or type produced. The material also must be confirmed as complying with the ASME specification by a certified material test report or certification from the material manufacturer (Subarticle NA/NCA-1220). Containment penetration materials meet the requirements of Articles NC-2000 or NE-2000 of the ASME Code, Section Ill, Division 1. The quality groups assigned to each component are given in Section 3.2. P:imipd The pressure-retaining materials are indicated in Table 6.11.

Materials for ASME Class i equipment are provided in Subsection 5.2.3.

Table 6.1-1 Engineered Safety Features Pnneipal Pressure-Retaining Materials l

[ Westiflgh0LIS8

' ' 'g a P

NRC REQUEST FOR ADDITIONAL INFORMATION ,

Question 25s.119 Table 6.1 1 in the SSAR lists materials for the ESF. However, this list lacks specificity, e.g., it lists 'sustenitic stainless steel." This list also refers to other sections of the SSAR v.'. re information may not be readily available.

For example, it lists the passive containment cooling system water storage tank in Section 3.8.4 of the SSAR. But the information on the materials cannot be found there. Further, this list may not be complete. Revise Table 6.1 1 '

in the SSAR to provide more specific information regarding materials used in the ESFs.

Response

SSAR Table 6.1-1 will be revised as follows to include the material information retuested: 1 SSAR Revision:

Table 6.1-1 '

Engineered Safety Features Principal Pressure-Retaining Materials Component Materials Core Makeup Tank Refer to Subsection 5.2.3 Passive Pesidual Heat Removal Heat Exchanger:

Tubing SM213TP304, Hemispherical Heads SAT 36lP304 Nozzles Sk336 F366 Tube Sheet $d3.i6iF364 In Containment Cooling System Water Storage Tank Refer to Subsection 3.8.3 Passive containment Cooling System Water Storage Tank Refer to Subsection 3.8.4.6.1 Spargers:

24 inch schedule 100 pipe @358 304Ef 316 8 inch schedule 100 pipe A 312 TP304.or jTP316 Pipe cap and Reducers 640LWPQoryP3J6 Containment Vessel and Penetrations Refer to Subsection 3.8.2 s y,g n,r- mrm py me#.m as y n.

m ,_ .a ma +u m s -- mw-am.

- 3 3.=- 7gygySugigei;555g3 252,119 1 l

W-Westingf ousel -

NRC REQUEST FOR ADDITIONALINFORMATION Question 252.126 Discuss hydrogen generation from the corrosion of materials within the comainment, such as aluminum and zinc, bas.ed on an assumed, jus'.ified corrosion rate (Section 6.1.1).

Response

ne hydrogen generation resulting from the corrosien of aluminum and zine within the containment are addressed in SSAR Subsection 6.2.5.

He third paragraph in SSAR Subsection 6.1.1.3 presently reads as follows:

• In post-accident situations or in situations ofinadvertent actuation where the containment is flooded with engineered safety features fluid, provisions are included to add solutions of sodium bydroxide to raise the pH of the fluid.

Section 6.3 describes the design and operation of the pit adjustment tank. Also, there is not a quantity of aluminum in the portion of the containment that can be flooded that would generate a significant amount of hydrogen.'

This paragraph will be replaced by the following:

SSAR Revision:

'In post-accident situations where the containment is flooded with water containing boric acid, sodium hydroxide solution is released from the pH adjustment tank in the passive core cooling system to the containment sump. The addition of sodium hydroxide to the sump solution is sufficient to raise the pH of the fluid to above 7.0. This pH is consistent with the guidance of NRC Branch Technical Position MTEB-6.1 for the protection of austenitic stainless steel from chloride induced stress corrosion cracking. Section 6.3 describes tiie oesign and operation of the pH adjustment tank.

In the post-accident environment, both aluminum and zine surfaces in the containment are subject to chemical attack resulting in the production of hydrogen. The non-ficated surfaces would be wetted by condensing steam but they would not be subjected to the boric acid or sodium hydroxide solutions since there is no containment spray. He bydrogen production analysis described in Subsection 6.2.5 includes hydrogen generation due to corrosion processes and conservatively assumes that all surfaces are exposed to the sump solution."

2s2.128 1 i

W westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION Uli 'iin

~

Question 252.134 Describe the materials of con *.uction for the major components of the demineralized water trea: ment system such as pumps, valves, and piping (Section 9.2.3).

Response

The demineralized wr :atment system is a nonsafety-related system. SSAR Subsection 9.2.3.2.2 of the SSAR contains the material scicction for the reverse osmosis manifold piping (316 stainless steel) and the degasifier piping, vahes, and fittings (stainless stect).

Material selection for other major components, such as pumns and valves, will be finalized during the equipment procuremeat design process. Material selection for all major components will be reviewed in accordance with the latest requirements. The material selectica will be based on current plant experience with the particular material and will incorporate lessons leamed as applicable to the component.

SSAR Revision: NONE 252.m W Westingtiouse l

NRC REQUEST FOR ADDITIONAL INFORMATION i ;f Question 252.135 Although the demineralized water system does not perform any safety related function, describe whether the design of the system ensures that failure of any ofits component would notjeopardize performance of the systems required for safe plant shutdown (Section 9.2.3).

Response

The systems required for safe plant shutdown do not rely on the demineralind water ystem to perform their functions. The systems required for safe plant shutdown are described in Section 7.4 of the AP600 SSAR.

SSAR Revision: NONE 252.135-1 V_/ Westinghouse

PiRC REQUEST FOR ADDITIONALINFORMATION i j Ouestion 252.140 Discusa provisions to address the potential for crosion/ corrosion of the steam and feedwater system. Justify that crosion/ corrosion will be insignificant for the projected 60-year plant design life (Section 10.3.6).

Response

The safety related portions of the main steam and feedwater systerm will be evaluated for erosion / corrosion using industry standard erosion / corrosion analyr,es methods. He design will ensure that the minimum wall thickness of the piping incuden erosion / corrosion allowance sufficient for 60 years of service at the design conditions.

SSAR Revision: NONE

[ WO5tiflgh00S8

i .

NRC REQUEST FOR ADDITIONALINFORMATION Question 252.144 Describe safety provisions that will te taken in the event of radioactive contaj instion of the fluids handled by the condensate polithirig system in order to twt the Al.A6t A raguirements (Section 10.4.6).

Resparce:

Radiological monitoring of the system provides for indication of sudden increar,ee in radioactivity above nornal levels. Upon exceeding pre-r.ct limits, automatic isolation of the steam generator blowdown sptem, which is the primary source of potential contamination, will be 3 Jtiated. Contaminated fluids can then tw processed thewgh the liquid radwaste system. periodic surveillance by the plant opera %rs of the radioactivity level within the system will provide indication of contamination within the system. The design includes provisions for installation of temporary shielding, and for handling and dispor,a! of radio.ctive fluids and resins.

SSAR Revision: NONE 252,144 1 l W Westinghouse i

1

NRC REQUEST FOR ADDITIONAL.INf0RMATION Question 260.1 Clarify /derme the neaning of the various terms that are uul to describe criteria for evaluating tuts rssults in Chapter 14 of the $$AR such as ' appropriate,*

• performance,' and ' acceptance."

Response' De following terms when used with regard to criteria for evaluating teat results have the following meanings.

Appropriate criteria: those criteria which are relevant to the test of interest.

Performance criterin: the criteria against which the succea or failure of ti;e teat will be judged. The dermition of ' performance criteria

• is derived fmm and ideatical to the definition of

'accep.2nca criteria

• in Regulatory Guide 1.68. Use of the wards 'jerformance criteria

• rather than ' acceptance criteria

• is to provide a distinction from the latter term at. it is used in the phrase ' Inspections, Tests, Analyes ard Acceptsace Criteria (ITAAC)*.

Acceptance enteria: ' Acceptance enterin' was not intended to be wed in the SSAR except when referring to ITAAC. ' Acceptance criteria' was unintentionally used in the following subsections of Chapter 14:

14,2.2.1 Con 6ct of Teat Program 14.2.8.1.67 Reactor Coolant System llot Functional Test 14.2.8.2.4 Reactor Systems Sampling for Fuelleading 14.2.8.2.26 Initial Criticality and bw Power Test Sequence 14.2.8.2.37 Power Ascension Test Sequence SSAR Revision:

' Acceptance criteria

• as used in these sections will k changed to ' performance criteria

• in a revisina to the SSAR.

260.M l W Westinghouse t

f

NRC REQUI.ST FOR ADDITIONAL. lNFORMATION ilm 'iH T l}

k-(

Quettion 260.2 he specific objectives of the initial test program stated in Section 14.2.1 of the SSAR only addreas the capability of 'he plant systems, stnactures, and cominnents. Section 1.1 of the SRP states that summary descriptions and FPWif6C objectives for each major pha5c of the test program lihould be provided. Section 14.2.l.2 of the SSAR states operating aaJ emergency procedures will be developed, tnal tested, and revised if necessary pnot to fuel loadmg. Section 14.2.5 of the SSAR also discusses the ur,e of plant operating and emergency pnwedates. Add a test program objective to validate pro;cdures dunny the conduct of the initial test program.

Response

A test program objxtive to validate plant operatmg and emergency procedures will be added to the SSAR as highhghted in the following escerpt from Sation 14.2:

GSAR Revision:

'14.2.1 Summary of Test Program and Objectives The purpn.c of this section is to dexribe the test program that is performed dunny imtial startup of the AP600 plant.

The escrall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform as required by the plant decign, and that activities culmmating in operation at full liceesed power, includmg initial fuel load, mitial enticahty, and power ascension, are perfonnut in a controlled and safe manner.

As required by 10 CFR 52.47 (a)(1)(vi), the inspections, tet.ts, analys.es and acceptance criteria relating to the AP600 design which are ne<.essary and sufficient to provide reasonable assurance that, if the inspections, tests and annipes are performed and the acceptance criteria met, a plant whkb references the design is built and will operate in accordance with the design certification for the AP600 may be found in the AP600 Inspections, Tests. Analyses and Acceptance Cnteria (!TAAC) Document.

De initial plant test program consists of a scrics of testa categonzed as construction and installation, preoperaticaal, and startup tests.

• Construction and installation tests are perfonned to determine that piant structures, components, and systems have been constructed or installed correctly and are operational. Some of these tests nay be part of the ITAAC program.

• Preoperational tests are performed after construction and installation tests, but pnor to initial fuel loading, to demonstrate the capability of plant sptems to meet performance requirements. Some of thew tests may be part of the ITAAC program.

I 280.2.,

w wesapause

I NRC REQUEST FOR ADDITIONAL. lNFORMATION

. mna I!!!" ]

• Startup tests, which begin with initial fuel loading, are perfcrined to demonstrate the capability of individual systems, as wc!! u integratt.d plant, to meet pe**onnance requirements.

The following are the specific objectives of the initial plant test program:

• Demonstrate that the plant construction is complete and acceptable.

• Demonstiste the capability of structurca, components arid systems to meet performance requirements.

• Demonstrate, where necest.ary , that the plant is capable of withstanding anticipated transients and postulated design basis events.

• Validate plant operating and emergency procedures as practical.

• Achieve initial fuel loading, initial criticality, and power avension in a controlled and safe manner.

• Bring the plant to rated capacity for sustained power operation.

Preoperational and/or startup teating is performed on those systems that:

• Are relied upon for safe shutdown and cooldown of the reactor plant under normal plant conditions and for maintaining the reactor in a safe conditinn for an extended Jiutdown peri (xl;

• Are relied upon for safe shutdown and cooldown of the reactor under transient and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown pericxl following such conditions;

• Are relied upon for establishing conformance with safety limits or limiting conditions for operation;

• Are classified as engineered safety features actuation systems (ESFAS) or are relied upon to support operation of engineered 6afety features actuation systems within design limits;

• Are assumed to function during an accident or for which credit is taken in the accident analysis and in the probabilistic risk avessment (PIL4); and

• Are used to process, store, control, or limit the release of radioactive material.'

260.2 2 [ W95tingt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION I

Question 260.3 Sc. tion 14.2.1.1 of the SSAR providea the purpose and scope of the construction and installation test program. This section also states that the test ab6 tracts will not be provided. State how the construction and ine.tallation tests will be developed and who will te responsible for performing those tests.

Response

Preparation of the detailed construction and installation tests are the responsibility of the constructor. During the course of construction, the constructor develops, performs and doeurt.ents various inspections, verifications, installation tests, cleaning, checkouts and other preparations w hich culminate with the energintion, irutial operation, adjustment and running of the equ pment. Development of the vnstruction and installation tests is bar,ed on the latest approved engineering information for the quipment to be uistalled and the system into which it is to be installed.

The SSAR will be nuxiified to reflect this reAponse as shown by the highlighted portion of the following excerpt from Section 14.2:

SSAR Revision:

14.2.1.1 Construction und int.tallation Test Propam Objectives The adequacy of construction, installation, and preliminary operation of components and systems is verified by a construction and installation test program, in this program, various electrical and mechanical tests are performa! using written procedures and instruc-tions, These tests normally include the following:

• Cleaning and flushing

• ilydructatic testing

• Checks of electrical wiring

• Valve testing

• Initial energintion and operation of eqitipment

• Initial calibration of instrumentation On a system basis, completion of this program will demonstrate that the system is ready for preoperational testing.

Abstracts for tests constituting the construction and installation teat program are not provided as part of this section. Preparation of the detailed construction and installation tests are the responsibility of the constructor. During the course of construction, the constructor develops, performs and documents various inspections, verifications, installation tests, cleaning, checkouts and other preparations which culminate with the energintion, initial operation, adjustment and running of the equipment. Development of the construction and installation tests is based on the latest approved engineering information for the equipment to te installed and the system into which it is to te installed.

~

280.3.,

W wesunmuse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestien 200.4 Section 14.2.1.2 of the SS AR states operating and emergency procedures will be developed, trial tuted, and revised if necceary prior to fuel loading. SRP Section 1.5 of the SRP also states that information regarding how and to what extent surveillance test procedures will be use-tuted should be provided. Addrna the development and trial-testing of maintenance and surveillance test procedures during the preoperational testing program.

Response

Plant operating, emergency, and surveillance procedures will be developed, trial tested, and revised if necest.ary prior to fuel loading. These procedures will be incorporated into the test program or otherwise verified through use to the extent practicable during the prwperational text program. Maintenance procedures are not required to be developed and trial tested by SRP Section 1.5.

The SSAR will be anodified to reflect this responw as shown by the highlighte i portion of the following excerpt from Section 14.2:

SSAR Revisiori:

14.2.1.2 Preoperational Test Prograrn Objectives Following construction and installation testing, proaperational tests are performed to demonstrate that equip-ment and systems perform in accordance with design ciiteria so that initial fuel loading, initial criticality, and subsequent power operation can be safely undertaken.

'lle general objectives of the preoperational test program are the folic. wing:

• Demonstrate that plant components and systenu, including alarms and indications, meet appropriate criteria based on design specifications.

• Provide documentation of the perfornunce and safety of equipment and systerra.

• Provide baselme test and operating data on equipment and systems for future use and reference.

• Operate new equipment for a sufficient period so that design, numufacturing, or installation defects can be detected and corrected.

• Demonstrate that plant systems operate on an integrated tesis Abstracts of preoperational tests are provided in Subsection 14.2.8.1. A listing is provided in Table 14.2 1.

W Westinghouse

NRC REQUEST FOR ADDITIONAL. lNFORM6ATION Plant operating, w,44mergency, and surveillance procedurea will te developed, trial-te4ted, and revised if necemry prior to fuel loadmg. 'Ihese prmcdures will be incorporatalinto the test progrem or olharwiw verified through pe, to the extent practicable, dusing the pieoperational test program.

All plant equipment us,ed in the performance of pro iperational tests will be operated in accordance with appropriate and wntten operating procedures, thereby giving the permanent plant staff an opportunity to gain practical experience in using ther,e procalutes and to demonstrate their adequacy prior to plant initit! criticality.

Several of the preo;wrational tests are performed during the plant hot functional test when the reactor coolant system is operated at operating temperature and pressure using the reactor coolac.t pumps as the heat source.

Subwetion 14.2.8.1.67 describes the hot functional test. Many other tests refer to this test for portions of the tests which require measurerrents at plant operating conditions.

230.4-2 W Westinghouse

NRC REQUEST FOR ADDITIONAL thf 0RMAT10N Ouestion 260.5 Section 14.2.2 of the SSAR states that test procedurca will be reviewed and approved by a startup coordinating group. Address the Regulatory Guide 1.68 guidance that the approved test pnwedures be made available to the NRC staff 60 days prior to their intended use.

Response

Copies of approved tcat procedures will be made available to NRC staff personnel from the Office of Inspection and Enforcement approsimately 60 days prior to the scheduled perfornance of the preoperational tests, and, for startup tests, not leu than 60 days prior to the scheduled fuel loading date.

The SSAR will be mmlified to refleet this respom.c as shown by the highlighted portion of the following excerpt from Section 14.2:

SSAR Revision:

14.2.2 Test Pncedures Preoperational and startup ints will be performed using detailed, step.by-step written pnwedures.

For each test, the test pucedure specifica the following:

• Objectives for performing the test

• P eriqu.Siten that must le completed before the test can be perfontwd

• Initial conditions under which the test is started

• 5:wial precautions arquired for the safety of per>or.ael or equipment

• Detailed step by-step instructions specifying how the test is to be performed

• Identification of the required data to be obtained and the methods for documentation

• Data analysis methods as appropriate

• Cnteria for test results evaluation Personnel with appropriate technical backgrounds and espenence will develop and review the pacedures using design information and recommendations regardmg test performance nyuirements and criteria provided by the major design organizations.

2co.s.,

w wesungnouse

NRC REQUEST FOR ADDITIONALINFORMATION Available infornation on operating and testing experiences of operating reactors will te factored into exh tes.t prmedure as sppropriate, I

Test procedures will be neviewed and approved by a startup coordinating group that willinclude owner / operator transgenent personnel.

Coples of approved tc6t procedures will be snade available to NRC staff perwnnel frorn the Office of Inspection and Enforcement approxinately 60 days prior to the scheduled performance of the prmperational tests, and, for startup tests, not less than 60 days prior to the scheduled fuel loading date.

260.5 2 [ Westingh0tlSe F

NRC REQUEST FOR ADDITIONALiNFORMATION h- O..!.

Question 200.0 Section 14.2.4 of the SSAR addresses the Westinghouse experience in the design, startup, and operation of more than 50 pressurized water reactor plants in the development of the initial preoperational and startup test program for the Ap600 plant.Section I.4 of the SRP states that a review of operating and tc6t experience at other reactor facihties and their effect on the test program abould be provided. The review should not be limited to only Westinghouse reactors, other sources of experience 6hould be reviewed in developing the initial preoperational and startup test program for the AP600 plant. Descrite how other sources of experience were considered in the initial preoperational and r.tartup test progrem for the AP600 plant.

Response

Other sources of experience reported and descrited in various documents such as NRC reports including Inspection and Enf ortement bulletins, and Institute of Nuclear Power Operations (IN PO) r ports including Significant Operating Event Reports (SOER) will also be utilized in the AP600 initial preoperational and startup test program. He AP600 design team experience with design tests will also te incorporated.

He SSAR will be modified to reflect this response ar. shon by the highlighted portion of the following excerpt from Section 14.2:

JSAR Revision:

14.2.4 Utilization of Reactor Operating and Testing Expenence in the Developnent of Test Program Westinghouse experience in the design, startup and operation of more than 50 pressurized water reactor plants is hz ' n utilized in the development of the initial preoperational a ' tartup test program for the AP600 plant. Other murrea of experience reported and described in various documer. . such as NRC reports including laspection an' ~.'nforcement bulletins and Institute of Nuclear Power Operations (INPO) reports including Significant Operating Event Reports (SOER) are also utilizalin the AP600 initial preoperational and startup tett program. The AP600 design team experience wi'h design tests is also incorporated.

Because of the standardization of the AP600 design, certain tests (designated as first of-a kind tests) will not be required on follow plants based on the experience gained during the startup of the fir-t, or lead, plant. Hese first-of-a-kind tests are identified in the individual test descriptions. (See Subsection 14.2.8).

W Westinghouse

l NRC REQUEST FOR ADDITIONALINFORMATION fpjMi:lii th ih Ouestion 260.7 Section 14.2.6.1 of the SS AR lists the minimum conditions for initir.1 core k= ding. Section 11.6 of the SRP states that precautions, prerequisites, and measurca consistent with the guidelines and regulatory positions contained in RO 1.68 should te included for initial fuel kuding and initial criticality. Add the appropriate prerequisites to the minimum conditions listed in Section 14.2.6.1

Response

he minimum conditions for initial core k>ading will ir:lude the following. He SSAR will be modified to reflect the additions as shown by the highlighted portion of the following etccrpt from Section 14.2:

SSAR Revision:

14.2.6.1 Initial Fuel lauding The minimum conditions for initial core lerding include:

• The composition, duties, and energency pmcedure responsibilities of the fuel handling crew are establi6hed.

• Radiation monitors, nuclear instrunwntation, manual initiation, and other devices to actuate building evacuation alarm and ventilation control are tested and verified to be operable.

• ne status of systems required for fuel loading is establistnl and verified.

• Re status of protection systems, interlocks, nnie switch, alarms, and radiation protection equipment is established for fuel loading and verified.

• Inspections of fuct and control rods have toen made.

• Containment integrity has been established and is being maintained under a surveillance program.

• De reactor veucl status han tua established for fuel hudmg. Components are serified to be in place or out of the vessel as required for fuel loading.

• Required fuel handling tools are available, operational, and calibrated, including indeting of the manipulator crane with a duminy fuel element. Dry runs have been successfully performed.

• Reactor coolant water quality requirements are established and the reactor coolant water quality is verified.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMAliON l119 liii

_ t

• Re res. tor veswl is filled with water to a level approximately equal to the center of the venel outlet nonjes. He water is circulating at a rate which provides reasonable usurance of uniform mixing.

• Re toron concentration in the reactor coolant is verified to be equal to or greater than required by the plant txhnical rpecifications for refueling and is being rnaintained under a surveillance program.

5 All sources of unborated water to the reactor coolant system have tan isolated and are urider a surveillance program.

• ne status of the core nakeup tanks and accumulators is verified as available and operable.

• At least two neutron detectors are calibrated, operable and located in such a way that changes in core reac-tavity can be detected and recorded. One detector is connected to an audible count rate indicator and a containngnt aacuation alarm.

• A resp (mse check of nuclear instruments to a neutron source is required within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rior to loading (or resumption ofloading,if delayed for B hours or nore).

Fuel aswmblies together with inserted componenta (control rods, burnable poison aswmblies, primary and secondary neutron sources) are placed in the reactor veswl one at a time, according to an established and approved sequence.

During and followmg the insertion of each fuel assembly, until the last fuel usembly has been loaded, the response of the neutron detectors is observed and compared with previous fuel loading data or calculations to verify that the observed changes in core reactivity are as expected. Sgaific instructions are provided if unexpected changes in reactivity are cbserved.

Because of the unique conditions exit, ting during imtial fuel loading, temporary neutron detectors may be used in the reactor vessel to provide additional reactivity monitoring. Credit for the use of temporary detectors nuy be taken in meeting technica' specifications requirements on the number of operable source range channels, 260.7-2 3 Westingt10USC

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 260.8 Section 14.2.6.3 ef the SSAR hsts tests that will be perforrral for the Erst AP600 plant only. Deae teata should be performed in accordance with Regulatory Guide (RO) 1.68. Wutinghouse should delete the 6rst plant only exception or provide justi6 cation for the execption to RO 1.68.

Response

In accordance with Regulatory Guide 1.68, SSAR Section 14.2 identifies tuts which verify that each AP600 is constructed in accordance with the Design Certification. In addition to those required by Regulatory Guide 1.68, Section 14.2 identifies 9 preoperational tests and 8 startup tuts to be performed only on the 6rst AP600 plant.

neue integral plant tuts are performed to verify design and analysis auumptions and to confirm design and analysis predictions. Becau e the AP600 is a standardire>d design, only construction verification tests are required on subsequent units. We 17 firs: plant mly tests neal not be perfortral on subsequent units.

He SSAR will be modified to reflect this rnponse as shown by the highlighted portion of the following excerpt from Section 14.2:

SSAR Revision:

14.2.1 Summary of Test Program and Objectives ne purpoi.e of this section is to describe the test program that is performed during initial startup of the AP600 plant.

He overall objective of me test program is to demcmstrate that the plant has been constructed as designed, that the systems perform as required by the plant design, and that activitica culminating in operation at full licensed power, includmg iniiial fuel load, initial criticality, and power ascension are performed in a cortrolled and safe manner.

As required by 10 CFR 52.47 (C(l)(si), the inspections, tests, analyses and acceptance criteria relating to the AP600 design which are necessary and suf6cient to provide reaumable assurance that, if the inspections, tests and analyses are perfo med and the acceptance criteria met, a plant which references the design is built and will operate in accordance with the design certif cation for the AP600 may be found in the AP600 luspections. Tests, Analyses and Acceptance Criteria (ITAAC) Document.

The initial plant tut program consists r Ie. series of tests catq.oriral as construction and installation, preop-erstional, and startup tests.

• Construction and installation rat., are perfornal to determine that plant structures, components, and systems have been coratructwi or installed correctly and are operational. Some of these tests may te part of the ITAAC program.

[ Westingh0USS

i NRC REQUEST FOR ADDITIONAL INFORMATION j

• Preoperational tests are performed after construction and installation tests, but prior to initini fuel loading, to demonstrate the espability of plant systems to neet performance requirements. Some of these tests nuy be part of the ITAAC program.

• Startup tests, which begin with initial fuel loading, are performed to demonstrate the capability of ,

individual systems, as well as integrated plant, to meet performance requirenwnta.

The following are th e specific objectives of the initial plant test program: {

t

• Demonstrate that AP600 design features meet performance criteria.- In addition, for the first AP600 plant, 9 preoperational tests and 8 startup tests are included to demonstrate that the AP600 performs as designed. 7 Design and analysis assumptions are verified and deelga and analysis pratictions are coafirmed. Because '?

of the standardized AP600 design, it is not necessary to repeat these tests during the initial test programs for successive AP600 plants. The other tests identified herein are adequate to verify that each subsequent AP600 has been constructed properly, There is no neal to reconfirm the design and analysis assumptions -

for standarized plants.  !

• Demonstrate that the plant construction is complete and acceptable.

• Demonstrate the capability of structurea, components and systems to meet performance requirements.

• Demonstrate, where necessary, that the plant is capable of withstanding anticipated transients and post. dated  ;

design basis events.

• Achieve initial fuel loading, initial criticality, and power ascension in a controlled and safe manner.-

• Bring the plant to rated capacity for sustained power operation.  ;

Preoperational and/or startup testing is performed on those systems that:

• Are relied upon for safe shutdown and cooldown of the reactor plant under normal plant conditions and ,

for maintaining the reactor in a safe condition for an extended shutdows period;

• Are relied upon for safe shutdown and cooldown of the reactor under transient and postulated accident c<mditions and for maintaining the reactor in a safe condition for an extended shutdown period following -

such conditions;

• Are relied upon for establishing conformance with safety limits or limiting conditions for operation;-

• Arc classified as enginected safety features actuation systems (ESFAS) or are relied upon to support operation of engineered safety features actuation systems within design limits; 260.8 2 W W6Stingh0t!S8 '

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• -wa m ~w *=e---- e--- * ~ = - - - - *'w =w

NRC REQUEST FOR ADDITIONALINFORMATION

• A.re anutred to function during an accident or for which credit is taken in the accident analysis and in the probabilistic risk ancament (PRA); and

• Are used to process, store, control, or limit the relene of radioactive material.

2m.8-3 W westinghouse

o ,

NRC REQUEST FOR ADDITIONAL INFORMATION ilF ili

-n Ouestion 260 9 Section 14.2.7 cf the SSAR stat.* that the schedule for fuel load and for each major plac of the initial test program will be provided by the owner / operator. While the specific dates will be supplied by the owner / operator, Westinghouse should incorporate minimum times for preoperational and startu,n testing specified in RO 1.68 or justify the exceptions.

Response

Because the AP600 design has fewer components and systuns and is not as complex as current plants, the schedule for conducting the preoperational phase and the initial startup phase may require less than the minimum timea of 9 months and 3 rnonths, respectively, specified in RO 1,68. As stated in the SS AR, the schedule for the initial fuel had and for exh major phase of the initial test program will be provided by the owner / operator in conjunction with the COL spplication. Any escephon to RO 1.68, if required, will be justified at that time.

SSAR Revision: NONE 2cos.,

w weeneouSe

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NRC REQUEST FOR ADDITIONAL.INFORMATION Question 260.10 l

Section 14 2.8 of the SSAR states the test abstracts provided identify what systems /comjonents will te tested tad what minimum information will be verified. In acwrdance with Section 11.8 of the SRP, Wutinghouse shauld also state that if the method of testing a structure, system, or component will not subject the item of system to repreuntative design operating conditions, the test abstract should contain sufficient information tojustify the test method to be used.

In addition to the five startup tuts listed in Section 14.2.6.3 of the SS AR as 'first AP60J piant only' there are nine preoperational tests and three additional startup tests that are listed as 'First Plant only.' Westinghouse should delete the words 'First Plant Only' from the '.ut abstruct titles of Chapter 14 of the SSAR. He owner / operator should justify why it will not perform specific tuts.

Response

If the method of testing a structure, system, or comsment will not subject the item or system to representative design operating conditions, the test abstracts contain sufficient information to justify the test nethod to le used.

He SSAR will be modified to reflect this ressmse as shown by the highhghted portion of the following excerpt from Section 14.2. For the response to the question regardmg 'first plant only' tests, sec Question 260,8.

SSAR Revision:

14.2.8 Individual Test Descriptions The test abstracts provided hereinafter idenufy what systems /commments will be tested and what minimum information will be verified. If the method of testing a structure, system, or component will not subject the item or system to representative design operating conditions, the test abstracts contain sufficient icformation tajustify the test method to te uvd.

[ WBStiflgh0US0

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 270.1 in Section 3.11.1.2 of the SSAR,

• Definition of Environmental Conditions,' where pi$tulatod high<rergy line failures are considered, a high-energy kne is defmed as a line with nominal diameter grc4ter than one inch. This defiration ts not consistent with the Standard Review Plan (SRP). Appendis A of Section 3.6.1 of the SRP (f1 ranch Technical Position ASB 31) defmes highenergy fluid systems u fluid rystems that, during nortral plant conditions, are either in operation or rnaintamed pressurimi under conditions where either the maximum temperature exceed 200'F or operating pressure excwds 275 psig, in accordance with 10 CFR 50.49, electrical equipment to be qualified includes equipment tlat is rehed upon to remain functional during and following design basis events, it is the staff's paition tlat design basis nents include high-energy systems as defined in Branch Technical Position ASil 31. Therefore, the definition of a high energy hne in the Ap600 SSAk should be change and made to be consistent with the SRP.

Response

The defmitmn of high energy lines is contained in SS AR Subsection 3.6.1.1, not in Subsection 3.11.1.2. The break stres evaluated are discunod in SSAR Subsection 3,6.2.1.2. To avoid the appearance of a definition of a high-energy line in Subwction 3.11.1.2, the parenthesis around

• greater than one inch

• will be rennved.

The revised sentence will read:

SSAR Revisiorc Postulated high energy line f ailures (as defined m Submtion 3.6.2.1.2) are assunxd in areas where high-energy lines prcater than one inch are routed.

2 7 0. , .,

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NRC REQUEST FOR ADDITIONALINFORMATION N

Ouestion 270.2 Section 3.11.2.1 of the SSAR staus that t,he turtladology for environmectal quahfication of electrial equipment is lawd on guidelmes providtd in IEEE standard 323 1983. To date, the NRC staff has not endorud IEEE 323-1983; therefore, references to this standard in its entirely or in part is not acceptable. As indicated in a footnote to 10 Cl R 50.49 and stated in NURE04588 and Regulatory Guide iJ9, the guidaxe in IEEE s.tandard 323-1974 is acteptable to the NRC staff for qualifying equiprnent within the scope of 10 CFR 50.49.

Response

The following will be added to the first paragraph of SSAR Subsection 3.11.2.1:

SSAR Revision:

As noted in the forward to tb standard, the 1983 revision of IEEE 323 1974 clarified the requirernents of the standard and gwat no additional requirements for qualifying Claw IE equiprnent. Therefort, qualification to the 1983 revision is equivalent to qualification to the 1974 revision.

i 1

W25tingh0USB

NFIC flEQUEST F0FI ADDITIONAL INIORMATION Question 270.3 In Section 3.11.2.1 of the $$AR, qualification by analysis is considered to be an eqitable orthod for environmethily qushfying electrical equpment important to r.afe-ty for the Ap600. Ihmever, in anordance with to CFR 50.49(f) s,nd NUREUMf 8, paragraphs 2.l(2) and 2.l(4), and in arco: dance with pievious NRC staff practice, qualificatim by analysis only is not nueptable. Therefore, environmental qualification of electrical equipnent important to r.afety for the AP6b0 6hould be in accordance with the requirementa of 10 CFR 50.49(f).

Response

In of der to clanfy our intent, the third and fourth paragraphs of Subustion 3.!!.2.1 willle combined as follows:

SSAR Revision:

When reliable data and proven analytical nrthmis are available, environmental quahfication may l ? bawd on analysis supported by partial type test data. This method includes justification of the theories and assumptionn und . . . . . . .

270.34 W Westlaghouse

• i NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 281,1 Table 5.2 2 of the SS AR lists ' recommended' reactor civ>'ent sptem (RCS) weer chemistry specifications. Specify the actual RCS water chemistry.

Response

Table 5.2-2 is the actual AP600 reactor coolant chemistry requirvments, lherefore, the word ' recommended

• will be deleted from the title.

SSAR Revision:

The SSAR will be modified to delete the word ' recommended

• from the title of table 5.2 2.

281.1-1 W WC5tiligh0tlS8

NRC REQUEST FOR ADDITIONAL.INFORMATION i

Question 281.10 Describe wby the guidelines in Sxtion 9.2.3.1.2 of the $$AR for demineralized water do not include specifications for halogens and sulfate.

Response

Table 5.2-2 in the SSAR whir.h provides limitations on the RCS water chemistry includes a halogen specification, in addition, the specification for specific conductivity in $$AR Table 9.2.31 is a de facto limitation on strong acid anions such as chlorides, fluorides, and sulfates.

SSAR Revisjon: NONE i

281.10 1

! W Wes1lnghouse

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NRC REQUEST FOR ADDITIONALINFORMATION ji' l!.i !

e.

Ouestion 281.14 Desenbe the runtimum stearn genenstor tube leak that can be s:commodated by the chemical and volume control system (CVC3) nakeup pumps (Section 9.3.6.1.2.2).

Pesponse:

The chemicrj and volume control system is designal to nake up for leaks up to 3/8-inch inside diameter instrument line break or equivalent steam gener6 tor tube leakage. With leakage up to this annunt (-120 gym) the trakeup pumps are sufficient to maintain teactor coolant system pressure and level above the safety iniection setpoints, in

(

the event of such leakage, this chemical and volume control system makeup capability permits the operator to bnng the plant to cold shutdown without using enginected safeguard systems. For larger leaks the engineered safeguards systems would be automatically actuated to naintain the plant in a safe condition. The chemical and volume control system makeup capability is not required or auumed in Chapter 15 safety analywa for these larger leaks but contributes to nakeup reliability in the PRA, SSAR Revision: NONE 281.w 1 W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 281.15 Identify the lea;4 tion where the hydrotest pump will te attached to the CVCS and discuss provisions to ensure that the systern will withstand tb pressure generated by this pump (Section 9.3.6.1.2.5).

Response

Flangxi connections are providal on the suction and dischstge header of the chemical and volume control system rnakeup pumps. These connections are shown on the chemical and volume control system piping snd instrumentation diagram (Figure 9.3.6.2). The pirtions of the chemical snd volutne control system which are uwJ for hydrotesting the reactor coolant system are designed to willutand the hydrotest pressure. The requirements for the hydrotest pump are to provide a flow rate of 35 gpm at 3200 psig. Also, temporary relief valves are used to ensure that the hydrotest pump does not overpressurire the system.

SSAR Revision: NONE

[ WC5tingh0Use

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NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 21,1,10 Provide a description of the mixed bed and cation tal demineraliurs. In addition, discuss provisions for spent resin regeneration (Section 9.3.6.2.1.1).

Response

Suhections 6.3.6.3.4 and 9.3.6.3.5 of the $5AR contain a ree detalini description of the catitm and mital tal demineraliurs. Alui, the irreartant equiprrrat parameters for the d:mineraliurs are given in Table 9.3.6-7. The demineraliurn uw dispauble resins. There are no provjeions for resin regeneration.

SSAR Revision: NONE 281.1 & 1 W Westinghouse l

NRC REQUEST FOR ADDITIONAL. INFORMATION RE Question 281.17 Describe the safety precautions for 6mring dw bydrogen um! for oxygen control in the reactor coolant (Section 9.3.6.2.4).

Response

The hydrogen tulles which are umi for otygen wotrol are stored in e location on the site that in at least 50 feet from any twildmg ccotaining safety-relatal equipment. Only a single bottle is connected at a tine. his linuts the hydrogen concentration following a leak or break in the ime.

SSAR Revision: NONE l

[ 281.17 1 l W Westinghouse l

i 1

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 281.18 Does the safety analysis of the plant takes credit for the injection flow produced by the CVCS makeup pumps during en accident? If credit is taken for this injection, die CVCS should be considered a safety re;4ted system and this would con'undict the defanititm of the system made ir. Section 9.3.6.1.1 of the SSAR (Section 9.3.6.2.4).

Response

The safety analysis d<ns riot t&ie credit for injection from the chemical and voluce control system makeup pumps in analyses where the actuation of the chemical and volume control system nakeup could adverxly affect a postulated accident, the analysis includes injection flow from the chemical and volume control system.

SSAR Revision: NONE 281.18 1 W WB5tingt10USB

NRC FIEQUEST FOR ADDITIONAL INFORMATIV Question 281.19 Explain why the hydsogai surply line, which is directly wnnected to the reactor ca,lant water purification loop in the CVC$ and penetrates the scactor containment toundary, has only one isolation valve. It should have two isolatier valves such as in the leadown and makeup lines in the CVCS. as it is required by General Design Criterion 55 (Sectmn 9.3.6 3.7).

llesponse:

A check valve (V094) is provided inside containment on the hydrogen addition line ar.d senen as a containment imlation valve. Airoperated valve V092 is outside containment on the hydrogen addition line and receives a containment swilatior, signal to provide containment imtation of this line outside containment. The use of a simple chak valve for isolation inside containment meets the contair. ment i6olation criteria specified in General Design Crit:rion $$. These valves are shown on the chemical and vol~ ne control system piping and instrumentation diagram (Figure 9.3.6.2).

SSAR Revision: NONE

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1

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestic 1410.2 Identify potential sources of esternal floca %g from components which at athin the AP600 design scope (Section 3.4.1).

Response

ne components which may be potential sources for external floodin, msafety-relatec', nonseismic tankt as shown in SSAL Fiyre 1.2-2:

• Fire water tanks as descrital in SuSsection 9.2.1. nese two tanks have volumes of 350,000 and 425,000 gallora, and are remote from safety-related structures.

• Condensate storage tank an described in Subsection 9.2.4. This tank has a volume of 300,s 3 gallons, and is located adjacent to the Containment and Auxiliary buildings. Tht.re are no doors or tsjor openings in these buildings at grade level in the vicinity of the tank.

• Demineralized water tank as described in Subsection 9.2.4. His mk has a volume of 150.000 gallers and is located adjacent to the Annex 11 building at elevation 107'-2*. Water will drain from the tank away t' rom the Annex 11 building to elevation 100'-0*. Nearby doors lead to areas in the Annex II building which do not contain 6afety-related components or systems.

- Boric acid storage tank as described in Subsection 9.3.6. This tank has a volume of 62,u00 gallons and 4 located adjacent to the demineralized water storage tank.

• .9iesel fue.i oil tanh as described in Subsection 9.5.4. These two tanks have volutaes of 100,000 gallons each. They are located remote from safety-related structures and are provided with dykes to retain leaks and spills.

Ir, addition, failure of the cooling tower or the service water or circulating water piping under the yard could result in a potent' d %a v urce. Ilowever, these potential sources are located far from safety-related structures and the consequence : e Aure in the yard would be enveloped by the analysis described in SSAR Subsection 10.4.5.

• lhe cooling tow.- basin u. telow grade and failure would act result in a potential flood source.

For the AP600, the ifW' building floor elevations are slighuy above the grade elevation, in addition, the slope of the yard grade directs water away from the buildings.

SSAR Revision: NONd 410.2-1 W Westinctiouse =

NRC REQUEST FOP. ADDITIONAL INFORMATION Round: 0 Q.Jestion Set: 09/23/92 i W II i Question 410.10 Tv/o lines are routed from the IRWST to each of the PXS compartments. The six inch line is routed to PXS-A and the 10-inch line is routed to PXS-B. What is the purge of these lines? Why are these lines sized differently?

What is the effect if the PXS compartment overflows (Section 3.4.1)7

Response

As descril ' in Section 6.3 of the AP600 SSAR, the IRWST has two separate injection lines (A and B) and each connects tt, one of the direct vessel injection lines. Each direct vessel :njection line connects to the reactor vessel dowacomer annulus. Each IRWST injection line is routed through r.ae of the two PXS compartments and these hnes are used to provide redundant injection flow paths to provide injection from the IRWST to the reactor coolant system when required.

These two IRWST injection 1 nes are also used to provide connections for several other purposes. A containment recirculation line connects trom each of the two recirculation ump screens to an associated IRWST injection line, upstream of the IRWST injection line isolation vaivea. These connections provide redundant containment recirculation flow to the reactor coolant system by using the IRWST injection lines.

In addition, tRWST injection line A is used to provide water from the IRWST to the spent tuel pit cooling system pumps. The IRWST injection line B is used to provide a suction connection from the IRWST to the normal residual hes: amoval system pumps. These two connections provide flow paths for IRWST filling, purifiaaion, sampling, and ^ oling. The connection to IRWST injection line B for the normal residual heat removal system can also be used during accident scenarios to take suction from either the IRWST or the containment recirculation sump and suoply injection flow to the reactor coolant system once it has been depressurized to within the capability of the normal residual hcat removal system.

He IRWST injection line A piping diameter is 6 inches to provide the proper reacte coolant system injection flow characteristics, based on the injection flow rates required following reactor coolant system depressurization when IRWST injection flow initiates. The IRWST injection line B supplies suction to the normal residual heat removal system and the line is required to be 10 inches in diameter to provide the required net positive suction head for the normal residual heat removal system pumps. He IRWST injection line B may r,lso be used to take a suction from the containment recirculation sump through the associated containment recirculation linc. The size of the recirculation sump line that contains the two series, motor-operated isolation valves is 10 inches in diameter for the same reason. The 6-inch IRWST injection line B is adequate to supply the spent fuel pit cooling pumps.

W W8Stiflgt10US8

NRC REQUEST FOR ADD!TIONAL INFORMATION Round: O E..

li Question Set: 09/23/92

. t in the event of a pipe break or LOCA in either PXS compartment, the water drains by gravity to the containment sump thr~agh floor drains located in each compartment. The floodup level in the PXS compartment wuld be the same as tl.e floodup level in the reactor coolant systern compartment in this situation. Each PXS compartment drain line contains redundant backflow preventers so tiit.1 reverse flow does not occur through the drain lines. ,

The effect of completely floodirig either PXS compartment is described in Subsection 3.4.2.2.2.1 of the AP600 SSAR, page 3.44. The opening to these compartmenta is located in the floor at elevation 107'2' and both compartments have ev-bs that are 12 inches high, to prevent any water on the 107'2' floor from druirdng into the PXS compartments. Should t's flooding in either PXS compartment continue such that the water fills a compartment and overflows the curb, the water then drains onto the floor at elevation 107'2.* Once the water flows onto the floor, the water would drain to the reactor coolant system compartment via the vertical access turnel since there is no curb surrounding the entrance to the vertical access tunnel.

Remote PXS valves and instruments are not assumed to function after they are flooded. The accumulctos are unaffected by flooding. Since a compartment would only flood at the rate that the containemnt floods, the CMT valves would be actuated and the CMT would empty before the CMT valves flood. One con:ainment recirculation line with motor operated valves could be flooded before they are actuated and would be assumed to fail. There are three other recirculation flow paths, the check valve path in the flooded compartment and the two other paths in the unflooded compartment.

No SSAR change is required.

l t 410.10-2 W Westingliouse l

NRC REQUEST FOR ADDITIONAL.INFORMATION f

Question 420.1 Table 1.8-1, " Summary of AP600 Plant laterfaces with Remainder of Plant," of Section 1.8 of the SSAR identified three items in the 1&C area, which were classified as Non-Nuclear Safety and not an interface. The staff is concerned about the environmental-related interface, espechtly the electromagnetic interference (EMI) with the digital instrumentation and control system. Addreas interface requirements which could cause EMI or other environrmntal-related impact on safety-related systems. For example, the microwave tower near the site could cause serious interference with the I&C systems.

Response

The environmental interfaces for the Instrumentation and Control Area are contained within the AP600 design and are addressed by the appropriate system specification documents. Electromagnetic Interference (EMI) and Radio Frequency Interference (RFI) limits are specified in the AP600 System Specification Documents for both safety-related and nonsafety-related instrumentation and control equipment, and identified to interfacing groups by this means. This is sufficient to address fixed sources of EMI/RFI in the AP600, such as microwave towers near the sit. . Portable sources of EMI/RFI, such as personal radios, will be controlled by administrative means.

In addition, EMI/RFIis addressed by the cabinet design features discussed in Section 4.0.6 of WCAP-13391(NP),

'AP600 Instrumentation and Control Hardware Descnption", Rev 0, May 15,1992 (Reference 3 of SSAR Subsection 7.1.6).

SSAR Revision: NONE W Westinghouse C

NRC REQUEST FOR ADDITIONAL INFORMATION Question 420.3 Provide the Prota. tion and Safety Monitoring System Process block diagrams listed in Table 1.7-5 of Section 1.7 of t% SSAR.

Response

Three copies of the Protection and Safety Monitoring System Process block diagrams were provided with the AP600 SSAR on June,26,1992. A discussion of this qu;,stion with NRC representatives on October,6,1992 in Pittsburgh confirmed receipt of these Prtsess Bkd Diagrams.

SSAR Revi-ion. NONE 420.3 1 W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 435.2 Section 2.3.2 of Chapter 11 of Volume 111 of the EPRI Utility Requirements Document for passive plants discusses the three-tier concept to provide a basis for rationalization of the arrangement of the plant electric systems. EPRI states that this concept permits establishing clear distinction between these systems based on their respective functional and operational requirements. Explain why this three-tier terminology is not usal in the AP600 design.

Response

Although the three-tier terndnology is not explicitly used in the SSAR, the AP600 design is consistent with the arrangement of the onsite power distribution system per section 2.3.2 of chapter 11 of volume 111 of the EPRI Utility Requirements Document (URD).

The SSAR figures 8.3.1-1,8.3.2-1, 8.3.2-2, and 8.3.2-3 illustrate the three tier arrangement. ne first tier are the main ac buses ES3, 4, 5, 6, 7, and 8 in SSAR figure 8.3.1-1 Sheet 1 of 2. nese buses contains the ac power distnbution systems feeding non-safety loads required for unit operation as described in the UR.D section 2.3.2.

ne second tier sie the main ac buses ESI and ES2 in SSAR figure 8.3.1 1 Sheet 1 of 2. Rese buses provides the ac and de power distribution systems supplying power to permanent non-safety loads. These non-safety loads, due to their specific functions, remain operational most of the time or wh n the unit is shut down, ne third tier consists of the de and low voltage vital ac power distribution buses as shown in SSAR figures 8.3.1-1 Sheet 2 of 2, 8.3.21, and 8.3.2-2 which provides the power distribution system feedmg safety (Class IE) loads. The AP600 ciectrical power distribution is consistent wit the three-tier concept described in section 2.3.2 of chapter 1I of volume til of the URD.

SSAR Revision: NONE

[ W85tingh0l!Se

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 435.4 Figure 11.2-1 of the EPRI ALWR Utility Requirements Document shows a unit auxiliary transformer (UAT) to be installed as a spare which is equal in aim to the other UATs. Figure 8.3.1 1 (Sheet 1 of 2) of the SS AR does not show the installation of a spare transformer. Provide justification for not including it.

Response

The following provides an explanation of the AP600 unit auxiliary transformer implementation and its relation to the ALWR Utility Requirements Document. The ALWR Utility Requirements Document do not terres.ent Regulatory requirements.

" Die SSAR Figure 8.3.1 1, AC Power System Station One Line Diagram, illustrates the main generator and onsite ac power system design of the AP600 The normal feed to the onsite power system is through the two Unit Auxdiary Transformers (UAT) which satisfy the functional requirements of the total plant electrical loads. EPRI URD specified an installed spare UAT, which can be connected to the system in a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period as a replacement to a failed UAT. Tids requirement was strictly based on improving the plant availability and investment protection only. The details of the spare UAT and its installation and construction details have not yet been fmalized.

Configuration of the spare UAT, additional operating benefits, and simplicity of the system are presently under discussion betweet. the EPRI personnel and the AP600 design team. The configurstions being discussed provide similar functional capability to the URD requirement on failure of a UAT.

SSAR Revision: NONE 1

l l

435.4 1 WOStinghouse

NRC REQUEST FOR ADDielONAL INFORMATION Ouestion 435.6 Does the design of the offsite power system, including its protection schemes, permit appropriate periodic inspection and testing? Describe the inspections and testing plans.

Response

The design of the offsite power system is a site-specific issue. The inspection and testing plans are documented by the Combined License applicant. Additionally, see SSAR Subsection 8.2.2.5 and Interface Item 8.4 of SSAR Section 1.8.

SSAR Revision: NONE l

l

' 435'6

W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION i!!!! iiiii n: g;

. t Question 435.11 Figure 8.3.1-1 of the SSAR, ' AC Power System Station One Line Diagram,' shows 4160 volt ac switchgcar for the reactor coolant pump motors classified as Class 1E. Section 8.3.1.1.1 of the SSAR states that Class IE circuit breakers used are for the specific purpose of satisfying the safety-related requirement of these pumps. Discuss the safety-related requirement for these pumps and the rationale for the need for safety-related switchgear.

Response

The AP600 reactor coolant pump circuit breakers receive a signal to trip from the protection and safety monitoring system upon generation of the following safety signals:

- Core makeup tank actuation signa!

- First stage automatic depressurization initiation

- liigh pump bearing water temperature Only the trip function of the circuit breakers is safety related. The pumps are tripped on core makt:up tank actuation and first stage ADS initiation to support LOCA mitigation by precluding interaction of the reactor coolant pump pressure head with gravity injection makeup to the RCS from the core trakeup tanks. The reactor coolant pump trip function is part of the engineered safeguards response to design basis LOCA's and therefore is implernented with Class IE circuit breakers.

The pumps are also tripped on a high pump bearing water temperature to protect pump flow coastdown capability following the loss of compo tent cooling water flow to the reactor coolant pumps.

SSAR Revision: NONE Westiflgh0llSB

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f NRC REQUEST FOR ADDITIONAL.lNFORMATION -

i l-Question 435.14 Perfortn a grid stability analysis for the AP600 design. The results should show that loss of the largest single supply to the grid (km not result in the complete loss of offsite power. The analysis should also consider the loss, through -

'- e single event, of the largest capacity being supplied to the grid, removal of the largest load from the grid, or loss of the most critical transmission line.

%e pid stability analysis should consider failure modes that could result in frequency variations exceeding the maximum rate of change determined in the accident analysis for the loss of reactor coolant flow (see Paragraph -

lit.l.f of Section 8.2 of the SRP).

I Response: g h- ,

l= The criteria and documentation of the requested studies are site-specific and will be discussed further by the l L -l

CombMed License applicant. This analysis cannot be undertaken without details of the specific utility grid.

SSAR Subsections 8.2.2.2 and 8.2.2.4 and Interface Item 8.3 of Section 1.8 discuss the need to conduct these analyses by tue Combined License applicant.

1 SSAR Revision: NONE l

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 435.15 Derme the facility's operating limits (real and reactive power, voltage, frequency and other) and provide a brief description as to bow these limits will be established. Also, describe the operating procedures or other provisions for assuring that the facility will be operated within these limits.

Response

Several factors are used to define the limits for the real power, reactive power, voltage, and frequency. These include, for example, the equipment ratig/ limitations as defined and designed by the manufacturer, system stability limits, system voltage requirements, and station auxiliary operating limia. Because of these factors, the facility operating limits cannot be fully dermed until specific equipment for the electrical system is selected, he real power and reactive maximum power limits are selected such that the unit will be turbine limited, not generator limited, and can provide the necess.ary ac voltage support to transmit this power to the utility grid. The capability is depicted by the generator capability curves, known as Vee curves.

I

' The minimum limits are established for real and reactive power based on the specific design parameters of the machine and the stability limits of the unit with relation to other units connected to the grid as discussed in RAI 435.14 and SSAR Subsection 8.2.2.4.

The minimum and maximum voltage linuts of the unit are within i 5 % of the generator nominal voltage as defined by the ANSI standards. ne exact limits within this range are determined after the offsite power grid has been designed, the station auxiliary system and the main step-up transformer impedance finalized. These limits will be determined so that the unit will remain stable and not be subjecrul to under or over excitation.

The frequency limits arc determined by the turbine manufacturer to limit potential damage of the turbine due to excessive vibration of the unit due to resonance.

The plant operating procedures will define these limits once they have been determined. Also, protective relaying devices will be installed to maintam unit operation within the allowable limits.

For AP600, the ac system electrical operating limits is not a safety related issue since ac power is not required for safe shutdown.

SS AR Revision: NONE 435.15-1 l W~

Westlflgh0ljSe 1

NRC REQUEST FOR ADDITIONAL INFORMATION 1 ' Uli!

Question 435.18 Section 8.3.1.3.3 of the SSAR states that the allowable current carrying capacity of the cables is lawd on the insulation design temperature. The derating is based on the type of installation, the conductor and ambient temperature, the number of cables in a raceway, and the grtmping of the raceways. The method of calculating these derating factors is determined from the insulated Cables Engineers Asscx:iation publications and other applicable standards. Section 8.3.2.4.2 of the SSAR states that where spatial requirements between raceways of different separation groups are not met, fire barriers are installed. The current carrying capability of those cr' lea wi!! be reduced when they pass through a fire barrier and as such a further derating factor should be applied. Describe whether further derating of the cables will be incorporated for those cables which pass through a fire barrier.

Response

While calculating the allowable current carrying capacity of the cables, the derating of the cables will be considered when those cables pass through a fire barrier.

The following statement will be added in SSAR Subsection 8.3.1.3.3 A

• Cable Derating *-

SSAR Revision:

A further derating of the cables is applied for those cables which pass through a fire barrier.

3 Westingt10USB

NRC REO' JEST FOR ADDITIONAL INFORMATION Ouestion 435.19 Specify the assumptions used in sizing the Class IE batteries to include the simultaneous statting of all connected loads with the maximum inrush for the first minute, and full load operation for the remaining period.

Responi.e:

The governing factor for the AP600 Class 1E battery size is the eteady state loading condition. He steady state loads are requited to operate for a long period of time; O to 24 bra and 0 to 72 hrs, compared to 0-2 hours nornully considered for the conventional nuclear pirats. Maximum inrush current has been considered for the valves (MOVs & SOL) that are required to operate in the first minute. The duration for the inrush current is conservatively usumed as one minute. The rest of the loads are steady state loads.

SSAR 59 vision: NONE

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1 NRC REQUEST FOR ADDITIONALINFORMATION E 'litt in 1 i

Question 435.20 Indicate whether the rating of the Class IE battery will be approximately 25 percent greater (at the minimum expected operating emperature) than that required ta supply the load requirements.

Response

AP600 Class IE battery sizing is based on the method, /en in IEEE 485,1983. In accordance with this standard, an aging factor of 25 % has t>een provided over and above the minimum operating temperature factor. Therefore, the ratmg of the Class IE batteries will be 25% greater than that required to supply the load requirements at the minimum expected operating temperature.

SSAR Revision: NONE

[ W85tirigt100SO

NRC REQUEST FOR ADDITIONAL INFORMATION lin ' "i:

E }

Ouestion 435.23 Section 8.1.2.1.1.1 of the SSAR states thai there are four independent, Class IE,125 Vdc divisions (A, B, C, and D). Divisions A and D are each comprised of one battery bank, one switchboard, and one battery charger.

Divisions B and C are each comprised of two battery banks, two switchboards, and two battery chargers. The first bank in the four divisions, designated as the 24-hour battery bank, provides power to the loads required for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an event of loss of all ac power sources concurrent with a design basis accident. He second battery bank in Divnions B and C, designated as the 72-hour battery bank, is used for those loads requiring power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the same event. It further states that no load shedding or load management is needed to maintain power during the safety actuation periods. Provide the list of loads fed by each lattery and the rationale for the 24-hour batteries.

Response

Refer to the SSAR Tables 8.3.2-1,2,3 and 4 for the list of loads fed by each battery. The AP600 passive systerns are actuated with power from the Class IE de and UPS system. Once the passive safety system actuation is completal, there is no fwther requirement from the Class IE de and UPS system; the systems will continue to perform their functions. The passive safety systems will be automatically actuated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after loss of battery charging capability (i.e. ac power) to the Cla.ss IE de and UPS system. This assures that the passive safety systems are actuated regardless of which licensing design basis event is postulated to occur. He time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after loss of all ac power sources for autcmatic passive system actuation was selected based on consideration of:

- a long enough time to provide a high probability of recovering ac power and avoiding ADS actuation.

- a short enough time to not unduly increase battery capacity requirements.

SSAR Revision: NONE l

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435.21-1 l.

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NRC REQUEST FOR ADDITIONAL INFORMATION Question 435.22 Verify that the load profile for the de batteries includes all those de and uninterruptible ac loads which are required to be operational for post-accident nonitoring and plant control following licensing hasis events.

Response

Required post-accident monitoring loads are included in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery load profiles. The plant control Icmds required for licensing design basis events are included in the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery load profiles.

SSAR Revision: NONE W85tingh00S8

NRC REQUEST FOR ADDITIONALINFORMATION Question 435.23 Provide a desenption of the capability of the battery chargers to supply power to Class IE de buses to properly function and rernain stable upon the disconnection of the battery, include in the description any nxxtes of operation that would require battery disconnection, such as when applying an equalizing charge.

Response

The battery cl.argers are equipped with a special circuit so that in the event of an unscheduled disconnection of a battery due to a fuse failure or otherwise, the battery chargers will continue to operate in a stable manner. Except during the trarnition time from normal to spare battery, the AP600 design does not require disconnection of the batteries for any mode of operati on including battery equalizing axxle. Scheduled maintenance and testing of the batteries will be conducted after the spare battery is connected. The downstream loads will be specified to withstand continously the maximum equalizing voltage of 140V de in accordance with the recommendation of IEEE Std. 946-1985, Table 1.

SSAR Revision: NONE 435.23 1 W westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 435,00 Provide a listing, by voltage class, of the following for the containment electrical penetrations:

s. 12t ratings,

b. umximum predicted fault currents,

't

c. identification of maximizing faults,

d. protective equipment actpoints, and '

e, expected clearing times.

Provide a description of the physical arrangement used to conne-t the cables to the containment penetrations, e.g.,

connectors. .plices, or terminal blocks. Provide supportive documentation that the 9 physical interfaces are qualified

• to withstand s. LOCA or sam line break environment.

Response

The specific information requested regarding 12: rating, maximum predicted fault current, protective equipment setpoints and expected clearing time, use of connector and splices is not available at this time. The information will be available when the vender for the containment electrical penetration anemblies is selected, As described in Subsection 8.3.1.1.5 of the S'AR, the electrical penetration assemblies will comply with the requirements stipulated in IEEE Std. 317 1983 and as endorsed by RG 1,63, Rev. 3. Primary and tackvp protections for the penetration conductors will be provi&4 in accordance :with IEEE Std. 741 1990.- The connectors, splices or terminal blocks as required for the connections of cables to the containment electrical.

penetrations will be qual fied to withstand a LOCA or steam line break environment.

SSAR Revision: NONE T westinghouse a

9 e NRC REQUEST FOR ADDITIONAL INFORMATION Ouer' ion 440.1 Provide updated or revised topical reports on the AP600 test program, including WCAP-13277. "Scs u.

a.

Design, and Verification of the SPES 2, the Italian Experimental Facility Simulator of the APouv s '

and WCAP-13234, ' AP600 Long Term Cooling Test Specification.' An accurate representation o.

facility design, a scaling analysis reflecting that design, a detailed test matri4, and an analysit plan shou!J be provided in the reports.

b. Provide 'opical reports detailing planned testing for the long-term check valve testing program and the departure from nucleate boiling testing program. Indicate whether the

• biased-open* check valves will be tested and, if so, a topical report detailing the test specification should be provided. If the ' biased + pen

• check valves are not to te tested, provide a detailed explanation why such testing is not requiM.

c. Provide WCAP-12980, 'AP600 Passive Residual lleat Exchanger Test Final Report,' that is referenced m the SSAR.

Response

a. WCAP-13277 and 13234 will be updated and forwarded by January,1993.

b. A topical report detailing the planned testing for the *long-term

• check valve testing program has not yet been prepared. Westinghouse has initiated a review of existing utility information to assess check valve opening performance after being closed at high delta P for a long time, i.e. conditions similar to those which wouhl be experienced by the gravity drain check vases, in order to assure that the test program will address relevant factors.

• Biased-open* check valves will not be tested in this program because these valves are open and are not exposed to high differential pressure.

Information on DNR Testing will be prc. Jed in January 1993.

c. WCAP 12980, 'AP600 Passive Residual lleat Exchanger Test Final Report,* will be forwarded by January,1993.

SSAR Revision: NONE

[ WOStingl100SB

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 440.4 The test matrix in WCAP 13345, Rev. 2 indicates that the maximum pressure to be tested in the CMT facility is approximately 1500 psia. Discuss why this is adequate, in view of the fact that the CMT will be approximately at the normal primary system pressure of 2250 psia when the safety systems are actuated. If the upper limit for the tests hu been changed, this should be indicated in the updated test specification requested in Q440.2.

Response

The CMT test facility has been designed and is being fabricated to operate at 2250 psig. A revised CMT test specification which includes a revised test matrix will be forwarded to NRC by January.1993. (Refer to response to Questions 440.2 and 440.5)

SSAR Revisiori: NONE 440.4-1 W Westiflgt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.5 The test nuttbers referenced in Section 8.0 of Revision 2 to WCAP-13345,

• Test Operation *, do not correspond to those shown in i'6hle 8.1, *AP600 Core Makeup Tank Test Matrix." There also ap; ears to be a similar inconsistency between the tests referenced in the

• Comments

• column of Table 8.1 and th3 seat numbers hsted in the left-most column. These incon.istencies should be corrected in the updated Test specification requested in Q440.2.

Responso:

A revised CMT test specification will be forwarded to NRC a.s discussed in the response to RAI 440.4. Section 8.0 of the revised specification will be made consistent with the test matrix. (Refer to response to AAI 440.4)

SSARIPRA Revisic ns: None l

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NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 440.6 The test descriptions in Section 8.0 of Revision 2 to WCAP-13345 are general in scope, and do not include detailed tat procedures. In some cases, temperatures, pressures, liquid le"cls, ai.d other test facility conditions are not specified, not are detailed data acquisition procedures discussed. The updated Test Specification requested in Q440.2 should include sufficient detail on test methods, facility conditions, and data acquisition, including step-by-step procedures, for the staff to determine if an adequate range of data on component performance will be pro :idul.

Response

When Westinghouse prepares a test specification, the detailed test procedures are not included since they are the responsibility of the testing organization. The testing organization develops the test procedures which are then reviewed and approved by Westinghouse. The procedures are operator instructions on how to operate the facility to obtam the test conditions which are specified in the test specification. 71 updated test specification (WCAP-13345) will provide viequate information for the staff to determine the adequacy of the tests including: information on the facility initial conditions, data acquisition, range of conditions, expected data, and instrumentation and method of testing.

SSAR Revision: NONE 440.6-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION -

O Question 440.7 Describe in detail how depressurization of the simulated primary eystem will be accomplish 4 and the means by -

which depressurization will be controlled by the signals from the level instrumentation to be tested in the facility.

Divuss how the depressurization rate will be related to that espected in the AP600 when the ADS is actuated.

Response

The depressurization of the CMT steam / water reservoir will be accomplished usine evo or more in vidual vent I paths, each with an independent open/close valve. Each vent path will contain an onnce sized such that the CMT.

test facility can be depressurized at a rate similar to the AP600 plant. "llie orifice sizes will be verified by pre-operational testing of the test facility. The first vent will be opened by operator action based on observed CMT level and subsequent vent actuation will be by operator action based on the facility pressure and/or CMT level. The prototypic CMT heated RTIT level instrument will Dql be used to control the vent valve opening.

SSAR Revision: NONE I

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NRC REQUEST FOR ADDITIONALINFORMATION Question 440.0 The check valves dowstream of the CMT in the AW,00 are now designed to be ' biased open." Will the check valve in the test facility dowstream of the test article also be ' biased open*?

Response

The check valvea to be used in the CMT discharge line of the CMT test facility will not be 'bia. sed open" check valves. This type of check valve (special orientation tilt d!ak) will be used in the plant CMT discharge line in order to maximize the reliability of CMT delivery (i.e. the check valvea are normally open). In the CMT test facility trx>re readily obtainable " swing-disk" check valves can be used and will have no impact on CMT delivery characteristics.

SSAR Revision: NONE

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W85tingh0USB

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 440.9 In Section 7 of the Test Specification and in Table 6.1 of WCAP-13345. no rationale is given for the choice of the data acquisition sampling rates for the various istruments. While the 10 llz sampling rate chosen for pressure and level instruments appears to be adequate, there is insufficient justification for the use of slower sampling rates for some thermocouples. Sampling the 29 internal thermocouples at 1 Ilz may miss fluid temperature fluctuations that could occur, for exstr.ple, during rapid steam injectica (and possibly jetting) into the CMT, or during 6ystem depressurization. His behavior may be important to understanding fluid mixing or the development uf therrrad stratific' tion. Provide an explanation for tiie 6ampling rates chosen, with particular attention to :he phenomena and processes that the instruments will be used to monitor.

Response

Westinghouse concurs that a faster sampling rate for the 29 fluid thermocouples may be helpful for observation of steam jetting, and flashing during depressurization. Westinghouse will increase the sampling rate of these thermocouples to 10 times /second in order to better observe the CMT behavior.

SSAR Revision: NONE W85tiligt10USB

NRC REQUEST FOR ADDITIONAL INFORMATION I

Ouestion 440.10 While the tests in the matrix of WCAP-13345 (Table S.1) include individual experiments to s:udy condensation, revirculation, depressurization, and dreining behavior, there does not arPear to be any test that takes the CMT through the entire sequence of events titat would be expected to owur in the plant. The behavior of the CMT as it goes through the transitions that occur during such a sequence is of substantial interest. At least one test that captures the entire series of states and transitions expected in the CMT should be ine'aded in the test snatrix.

Response

As stated in the responses to Questions 440.2, 440.4, cad 440.5; WCAP.13345 will be revised and provided in January,1993. The revised test matrix willinclude testing that simulates all the CMT transitions in one test.

SSAR Revision: NONE W Westingtiouse

I NRC REQUEST FOR ADDITIONAL.INFORMATION 1!it "!!

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l Ouestion 440.11 Revision 0 of WCAP-13342, ' AP600 Automatic Depressurization System Test,' is dated January 1991. An updated version of this test specificatior, should be provided, incorporating any changes in the design or test plans for the test arti :les in Phue A and Phase B, particularly as a result of changes in the AP600 plant design.

Response

WCAP-13342, AP600 Automatic Depressurization System Test is currently being updated, specifically to incorporate additional or revised information for the Phase B of the test program. This revision will also include changes to the Phase A test program. The revised WCAP will be provided in January,1993.

SSAR Revision: NONE i

0.1 b1 W Westingtlouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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Ouestion 440.12 aince Revision 0 to WCAP-13342 was issued, the 4th stage of the ADS has been redesigned to incorporate larger velves (12' vs. 8*) and a different configuration with respect both to number of valves und connection to the RCS hot leg piping. Westinghouse has previously committed tc test these valves as part of the AOS test program; however, no information has been provided to the staff specifying the test loop configuration or the test nuttrix envisioned for these tests. His informrtion should be provided for staff review. The level of detail in this information should be commensurate with that requested for the Phase A and Phase B tests.

Response

The size of the fourth stage ADS valv. has increased from an 8-inch valve to a 12-inch valve. A specific test will be run on the founh stage valve. The fourth stage ADS valve test will examine not only the flow effects of the valve geometry, but the mechanical design aspects of the valve design such as opening torques, loads, sealing capabilities and sizing of the valve operator. He component testing of the valve will be performed as a component verification test following design cer'.ification, when the detailed valve design is complete.

In the interim, if there are questions on the flow through the fourth stage valve, they can be addressed with sensitivity studies. Ite resulting flow behavior will be confirmed by the fourth stage tests performed on the specific valve design which will be used in the plant.

SSAR Revision. NONE u .12-1 W westingtiwse

NRC REOUEST FOR ADDITIONALiMORMATION y .. y

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Question 440.15 ne opening tirnes for the ADS valvo dunng the Phav 11 tuts descnbod in WCAP-13342 are considerably shorter than the titan capated in the plant. This is deacnbed as being a ' conservative

• approach, in that snechanical loads due to air clearitag, water harnmer, etc. will be greater in the teat facihty than would be expected in the actual plant, llowever, the imtial depressuriation rate .n the tnts will alt,o le grnter than that in the plant, a reault that could be considered 'non<onsenstive.' Of additional concern is the fact that the rapid openir g in the test facility will not allow sufficient data to te collected on valve flow rates and system depreuu-intion brhavior as a function of time during the slow opening process to permit their characteriution for pinnt analysu. Discuss how the ADS valve behavior during opening will Se determined.

Responso:

The r alve tchavior will be established in the following numner (please refer to Table 8.2 of the 'AP600 ADS Test 5p a.ication', WCAP.13342).

• Test B01 nx;uires the ADS Stage 1 valve to open at full initial prnsure (2500 psig) and delta P. This test wili be conducted using saturated steam and the prototypic AP600 valve stroke time. his test accurately simulates ADS mitiation at a full system prusure.

• Test D02 thru D05 will be run with the Stage 1 ADS vahe at 1/4,1/2,3/4, and full open. These tests will le conducted using saturated steam, and will be used to simulate the single phase Dow vs. della P vs.

upsti arn pressure for each valve position. The existing facihty steam flow control valve and isolation valve are used to quickly initiate the steam now. Baad on the Phue A testing, the steam pressure upstream of the full open ADS Stege 1 valve that can be duplierted is 2100,1100 psig.

• Test B06 will demonstrate that the Stage 2 or 3 ADS valve will opca at an initial detta P as high as 2500 psig. This test will be conducted with saturatal steam. He Stage 2 and 3 ADS valve stroke time for this tut is much futer than the actual plant valve stroke time (10 nc. vs.90-120 sec.) in order to nutintain as high an opstream prenure and delta P as possible throughout e entire opening stroke. He valve will be stroked close after reaching the full open position.

• Tests D07 through DIO will be run with the Stage 2/3 valve at 1/4,1.'2,3/4. and full open fixed positions, respectively. These tests will be conducted using saturated steam, and will be used to dem<mstrate the single phne flow v6. delta P vs. upv.ieam prewee for each valve position. From this information the flow vs. valve strQe can be interpolated. The existi..g facility steam flow control valve and isolation valve are used to quickly initiate the steam flow. Based on the Phase A testing, the steam pressure upstream of the Stage 2/3 ADS valve that can be duplicated ranges from 2100 psig for the 1/4 open valve to 10n0 psig for the full open valve.

WO5110gt10USB

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1 NRC REQUl.ST FOR ADDITIONALINFORMATION i

lich of the valves will te tuted in a sirrular way in the subsequent portion of the Phase D tuts that use ' saturated' water frc n the bottom of the facility steam / water supply tanlr. %ese testa will clor.ely simulate the minimum void fraction condition that can occur at the valve inlet and produce the maximum rnasa flow. 'Ilat is, the water leaving the lottom of the supply tank contains a minimum void fraction and the water increases slightly in void fraction due l to the line lonwa. Scoping analysca for example, show that the maximum void fraction at the Stage i ADS valve )

inlet will be - 15 % at a matimum flowrote of ~290 lbs./sec., while the presturiur prasure will remain relatively constant (2300 psig to 2100 psig) throughout the open/close cycle test, D12. l The open/close cycle tut D17, for the Stage 2/3 ADS valve will be performed from full system pressure, ~ 2250 ,

psig, in o s to denumstrate the operability of the valve if both Stage 1 ADS valves failed to open and initiate plant depressuriration. Scoping analysis for this test indicate that the water void fraction at the supply tank exit and at the valve inlet are 45% and 50% respectively at a maximum expected mass flow of ~ 1200 lbs./sec.

Tests 1113.D16 and DIB 21 will be performod with the Stage 1 ADS valve and Stage 2/3 ADS velve respectively at 1/4,1/2,3/4, and full open, fixed positions. %ere tests will be performed to characterire the flow vs. della P vs. valve position at the minimum void fraction condition. For these testa, flow will tu initiated by opening the 12-inch valve in the prusunar discharge line. This valve will have a r eroke time of 1015 seconds no that full flow can le quickly achieved, maximir.ing the pressuriter pressu.e and available mau.

Throughout all the Phase D testa the valve performance will be continuously nxmitored using the MOVATS instrumertation system develnped and unal for testing valve operability at operating nuclear power plants. De MOV ATS instrumentation will include a digital data acquisition system that will provide imnwdiate post test data reduction for evaluation. Acquired data willinclude:

• upstream and down tream pressure nrasurements

• valve position vs. time e valve stem force e motor rpm and developed power This measurement system combined with the structured test matrix will obtain sufficient inforrration for the ADS valves with the flow vs. time providal by the 1300 ft' steam / water supply.

Post test evaluations will include valve disassembly and insitction.

SSAR Revision: NONE 440.15-2 W Westinghouse

NRC Hl:OUEST FOR ADDITIONALINFORMATION Question 440.19 he 'nornal opening conditions' for the stage 1, 2, and 3 ADS valves are given in WCAP-13342, Rev, O as approximately 2250 psia, 800 psis, and 300 psia, respectively (section 5.U). linwever, analyses of SDLOCAs in the APM0 performed by the NRC and the applicant show that each stage of the ADS nuy open over a very wide range of preuure, depending on the scenario, since ADS actuation is contrc!ied by CMT level. Discuss how the opening of the valves, either simulated (Phase A) or real (Phee B, 4th stage) during the ADS t .sts will be sequenced as a function of pressure to account adequately for the range of preuure over which each valve must operate.

Response

All the cycle open ADS valve teats in the Phuc B portion of the ADS test will te initiated at full RCS pressure, 2250-2500 psig. The steam / water supply pressure at the completion of the opening stroke will vary depending on the valve being testea (i.e. 4' globe vs. 8' gate valve), the preuurizer discharge u.sd (i.e. steam or saturated water), and i: c actual vahe stroke times. Ilowever, in all cases the entire opening stroke will be performed at snuch higher upstream valve pressures than currently prnlicted in the safety analysis.

The valve flow vs. delta P vs. position tests will aim be performed over the maximum pressure range within the tacility capabilities. The maximum expected pressures for the Stage i ADS valve will be ~2100 psig for the full open position and ~1000 psig for the Stage 2/3 valve at its full open position when saturated steam is being supplied to the valves. Of course, higher upstream pressures will be achieved when the valves are partially open or when saturated water is supplied to the valve::.

He comi,leted Phai,e A tests with saturated steam achieved the folbwing simulated ADS conditions:

• All three sJges open - the valve upstrearn preuure simulated was from 500 psig to 15 psig.

• Stage 1 open - the valve upstream ),reuure simulated was from 2l00 psig to 800 psig. (Stage I flow expa;ted during the opening stroke was also simulated)

• Stage I and 2 open - the valve upstream pressure simulated was from 10(k, psig to 15 psig.

These Phase A pressure rangea and resulting volumetric flowrates exceed the ranges analyzal in the Chapter 15 analysis.

SSAR Revision: NONE 440,19 1 W Westingh00SO

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NRC REOULST FOR ADDITIONAL INFORMATION Ouestion 450.4 Table 6.4-2 of the SSAR identifica several site chemicals. De SSAR states that analysis of their sources are in accordance with Rerulatory Guide 1.78, llowever, the $$AR does not addreas chlorine L aser, Table 1.91 of the SSAR states that Regulatory Guide (RO) 1.95, ' Protection of Nuclear Power Plan <mtrol Room Operators Against an Accident Chlorine Releau,' is not applicable to the AP600. Provide your rationale for not sadressing the chlonne gas exposure. If chlorine relene is considered site-specific, it shov'd be so identified and the future licensec/ applicant should be required to conform with RG 1.95 (Section 6.4).

Response

ne dc4ign of the demineralized water, waste water. and cooling water systems does not include the use of chlorine in the chemical treatment schemes for each of these systems. Herefore, no accidental release of chlorine could occur within these systerm. Other potential sources of chlorine gas not associated with the AP600 plant (such as railroad cars or truck tankers) are considered site specific and will be addreswd by the Combined License applicant.

SSAR Revision: NONE W Westingtiouse

NRC REQUEST FOR ADDITIONALINFORMATION I

Question 450.5 Discuss the elevation and location of plant vents, including positions relative to the control room ve-Wation inlet (Section 6.4).

Response

The control room outside air inlet is located approximately 115 feet laterally and 86 feet vertically telow the plant vent discharge point. He bottom of the control room air intae is approximately 59 feet almve ground elevation.

De location of the control room outside air inlet antisfies the SRP 6.4 requirement that the control room ventilation inlets should te separated from the major potential relene points by at least 100 feet laterally and 50 feet vertically, it further natisfies the RO l.95 guidance that the fresh air inlet should be at least 15 meters above grade elevation for protection from toxic gases.

SS AR Figure 1.21I shows the plant vent and air intake on a plan view, ne plant vent discharge point is near the interwtion of building column lines 4 and J 2, approximately at elevation 250 feet. Re control room outside air intake is near the interrecticm of building column lines 8 & J on the auxiliary building nef SSAR Figure 1,217 r, hows that the bottom of the control room outside air intake, approxima'ely at eleution 157 feet.

SSAR Revision: NONE

' 450.5 1 W Westingh0Use

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NRC REOUEST FOR ADDITIONAL.lNFORMATION Question 850.7 WCAP 13053 states that the MCR and contairunent of the AP600 do not have pet.t-ac<ident ESF astrospheric cleanup systems. Sxtion 6.5 of the SSAR ab.o states that ESF filter systenu are not applicable to the AP600 design. Ilowever, Section 6.4 of the $$AR credits the VBS when ac power is available to provide normal and abnormal llVAC scivicea to the MCR sad other associated areas. Section 9.6.1 of the SSAR states th supplemental air filtration subsystem is designed to meet RG !.140.

Provide justification for not conforming with the guidaa.ce of RO l.52 for the ESF atmospheric c canu}, system (supplemental air liltration subaystem of VBS) for the control room while crediting it during abnormal as well as normal operation when ac c W avaii**4e. Also, clarify your staternent in Section 6.51 of the SSAR whi:h t 4 iliuW. This is inconsistent with the credit for filtration by the VDS claina that the ESF filter i wm i -

during abnormal as well e. tui V :F +. ' control room and other casociated areas (Section 6.4).

Response

The AP600 MCR operator babitsbib' requirements under accideat condi; ions are piovided by the main control room emergency habitability system (VES). The VES is designed to satisfy nuclear safety re'ated systern design and seismic Category I requirements.

The nuclear island nonradioactive ventilation systern (VBS)is a nonsafety-related system except for the components that provide isolation of the mein control room envelope. The atmospheric cicanup systeci (supplemental air filtration subsystem of VBS) operating danng abnormal nodes is a defenw-in depth function when VBS is operabh and an oc power hource is available. There is no credit taken 'a VBS filtration operation in the MCR habitability analysis under accident conditions.

RG 1.52 applies only to the post accident engmected+afety-feature (ESP) atmospheric cleanup syn. tem and does not apply to atmospheric cleanup systems der,igned to cetlect airbome radioactne materials during normal operation, includmg anticipated operational transients. Therefore, the application of RG 1.140 for normal air filtration system design is appropriate.

SSAR Revision: NONE l

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NRC REQUEST FOR ADDITIONAL.INFORMATION K

Ouestion 460.2 Correct or clarify the following information obtained from the identified tables in the SSAR relating to secondary coolant concentratwn (Sections 11.1 and 11.2):

Tables 11.1-4 and 11.14: Total steam generator (SG) blowdown flowrate 4.2 x 10' lb/hr Table 11.24: Total 50 blowdown flowrote 8.4 x 10'lb/hr

Response

The total steam generater blowdown flow rate shown in SSAR Tables 11.14 and 11.17 (4.2 x 10'lb/hr)is used to calculate the wcondary cmlant concentrations for use in in-plant evaluations. The lower blowdown rate (comparix! to Section 11.2) tends to retain note radioactivity in the secondary coolant, resulting in conservative t,hieldmg, ALARA, and accident source terms. Because the intended use of the parameters in SSAR Section 11.2 is to estinate 'ffsite releases and determine if the radwaste systems are adexluate to maintain effluent releases to uncontrolled areas within die limits specified in 10 CFR 50, Appendix 1, the maximum c<mtinuous capacity of the syr. tem as shown in SSAR Table 11.24 (8.4 x 10' lb/hr) in u>ed.

SSAR Revision: NONE w weenpouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 460.7 Table 2 of Section 11.$ of the SRP, " Process and Efnuent Monitoring Instrumentation and Sampling Systems,'

includes a service water system effluent nonitor. De staff notea that AP600 design includes an upstream provision for this monitor in the form of component cooling water system nonitor, ne staff does not consider an upstream provision as an adequate basis for eliminating a downstream provision for this nonitor. Herefore, include a service water system tuon1 Lot or justify its elimination (Sections 11.$).

Response

he Service Water System water inventory is integruted with the Circulating Water System and the cooling tower basin. The Service Water Sy6 tem does not have a dedicated effluent discharge pathway but shares a common c4 cling tower blowdown system connection with the circulating water system as shown in Figure 10.4.51. As shown in Table 11.$.1, the normally nonradioactive systems which provide potentialleak pathways into the service water sy6 tem include radiation nonitors with nominal minimum detectable concentrations of 1.0E-B uCi/cc. Further dilution in the service water er circulatmg water systems or in the cooling water blowdown system, as described in Section 11.2.3.3, would reduce the concentration to approximately 3E Il which is below the minimum detect:ble concentration for a continuous-type monitor. Therefore, a radiation nwnitor in the cooling tower blowdown (service water efnuent) will not provide additional information or control of effluent releases.

As discussed in Section i1.5.3, the primary means of quantitatively evaluating the isotopic activities in ef 0uent paths is a progism of sampling and laboratory measurementa. Table 9.3.4 2, which indicates grab sample locations, identifica a grab sample for the cooling tower blowdown as well as the cooling tower basin and service water basin.

The grab sample allows for measuring of the activities in the ef0uent path.

SSAR Revision: NONE f

l 460.7 1 W Westinghouse 1

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 620.0 Explain the meaning of M Mis as used in the AP600 design. Exactly what is included in M Mis (Section 18.1,

p. 18.1 -1)7

Response

ne M MIS (Man Machine Inte face System) as used in the AP000 is the same as the M MIS used in the ALWR URD, Chapter 10. The AP600 M MIS includes instrumentation and control systems provided as part of an ALWR plant which perform the requisite monitoring, control and protection functions au.ociated wi:h all modes of plant operation (i.e., startup, shutdown, standby, power operation and refueling) as well as off-normal, emergency, and accident conditions. See the figure attached. The following paragraphs below provide the dermition and mission of the M MIS, and are consistent with those found in SSAR Section 18.!.8.1.

'The AP600 M-MIS is con, posed of those plant systerra that perfonn the mcmitoring, control, and protection functions in the plant. It also includes the nelection, synthemir.ing, and distribution of process data to other plant personnel who have use for it. These added users include, management, engineering, and traintenance personnel.

The M MIS scope includes not only the plant data acquisition and control systems, but also interfacing systems such as the switching and tagging sys tem, the health physics and chemistry data systems, the traditional technical support center (TSC), operations support center (OSC), and emergency offsite facihty (EOF) information systems, the computerized plant process procedures, the management plant information systems, the maintenance information system, and planning and scheduling. It includes designing to accommodate both the physical and the cognitive characteristics of humans involved in the use, maintenance, and control of the plant.'

'It is the mission of the AP600 M MIS to improve the means that are provided to the users of the plant operation and control centers for acquiring and understandmg plant data and in executing actions to control the plant's processes and equipment, This improves, the individual's reliability and team work.'

SSAR Revir, ion: NONE 620.64 V,f Westinghou e

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Question 620.7 What is meant by *... AP600 has a plant-wide network that provides pre-processed plant data...' (Section 18.1, p,18.1 1)?

Response

The AP600 usea a distributed computer architecture. Plant operating data is placed on the architecture's data highway for operations use, it may te useful to different users of the highway for different purposes and possibly in different forms. For instance, there are times when the ability to examine the raw plant data inputs would be useful information. Here are still other tirnea when it would clearly te more meaningful to te able to examine the results of a computation, and not merely the raw input data. Part of the task of the interface designer is to decide what plant information should te accessible by what users at what times and in what forms.

The distributed architecture c<mtains nodes where computations are performed and then redistributed across the data highway. Further explanation of this architecture can be found in subsection 7.1.1 of Chapter 7 of the AP600 SSAR This distributed highway is the ruonitor bus, shown in SSAR Figure 7.1 1.

SSAR Revision: NONE 620.7 1 W Westingh0USB

! NRC REQUEST CUR ADDITIONAL INFORMATION i Question 620.11 i

What constituica ' general literature' (Section 18.3, p.18.3 2)?

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Response

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• general literature,' Watinghouse means publicly available doeurnents describing researth results and design

principles and guidelines from relevant disciplines such as human factors, industrial engineering, and applied

p*ychology. Examples include articles in professional journals such as lluman Factors, laternational Journal of i Man Machine Studies, and lluman-Computer Interaction; conference proceedings such as the proceedings of the recent 1992 IEEE Fifth Conference on lluman Factors and Power Plants held in Monterey, Call books such as the .

j forthcoming edited volume Verification and Validation of Iluman Machine Systems, which covers the results of a j NATO conference on that topic; and technical reports from government sesearch organizations such as NASA, i Armstrong laboratory, the FAA and the NRC.

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! SSAR Revision: NONE i

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P s2om W westinghouse

NRC REQUEST F0FI ADDITIONAL.INFORMATION 1*H* inii Ouestion 020.18 What is n eant by a 'near full-scope, hi fidelity, simulator" (Section 18.5)?

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Response

SSAR Subsection 18.8.2.3.1.7 discurs.es a mod:1 of test ted fidelity that descritus the dimensions along which testbeds can vary in fidelity (See Figure 18.8.2. 10). The two major dimensions are realism (i.e., the degree to j

which the tentled looks and/or behaves like the actual system, and completeness (i.e., the scope of coverage of the teatted - how much of the total system is included in the teetbed). Realism is further broken down into physical l l

fidelity (the degree to wh;ch the testbed physically resembles the actual system hardware), and functional fidelity, which includes (1) tae degree to which the information content resemble 4 the infonnation content of the actual system (e.g., with respect to number and content of displays), and (2) the degree to which the teatt4 displays the same dynamic characteristics as the actual system. For each of the reventeen evaluations define'in the SSAP., we specify the minimum testbed fidelity requirements for conducting the evaluation. The specification includes requirements with respect to physical form, information content, and dynamics.

SSAR Table 18.5-2 provides an abbreviated summary 4 he t testbed requirements that are more fully specified in the seventeen individual evaluation descriptinns. Part Task Simulator refers to testbeds that may be limited with respect to completene6a of scope (e.g., the testbed may include the workstatior displays but not the wall panel information system), and with respect to the degree of dynamics displayed (e.g., static displays may be used).

Near Full-Scope liigh Fidelity Simulator refers to a control room simulation that is high on both the completeness and realism dimensions. A near full-scope high fidelity simulator would display high physical fidelity (i.e., the te6tted would physically resemble the actual hardware to le implemented in the AP600 control room), as wc!! as high fidelity s . i respect to information content (the number and content of displays), and underlying process dynamics (i.e., it would be driven by a high-fidelity AP600 plant simulation). The nxxlifier 'near* is umi to indicate that features of the simulation that are not relevant to the tests being mad; may not be full-fidelity. For example workstation displays covering portion of the plant that will not be acce4 sed in the tests may not be implerated in the test simulator.

SSAR Revision: NONE W Westinghouse i

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NRC REQUEST FOR ADDITIONALINFORMATION N. ., ~1 Question 620.22 Section 18.5 2 of the SSAR statea that ' Acceptance enteria related to plant construcuan will be the responsibility I of the Combined 1.iceme lioider.' Wht is meant by ' acceptance criteria related to plant construction' l (Section 18.5, p.18.5-2)7

Response

ne ITAAC descrilut in SSAR Table 18.8.2-2 are designated as teing either Design ITAAC of Construction ITAAC. The only constmetion ITAAC are thme that require the completed room design sad an actual training simulator level of fidelity to perform, that is, the conformance with human engineering guidelines and the validation of the M MIS. All of the other ITAAC listed within this table are considered to be design ITAAC and will be performed by the vendor. The wastruction ITAAC will nml to be performal once a training simulator and simulation of all interfaces to the control room can be simulated, and should therefore te performed by the combined operating licent.e holder. This is true as long as the COL holdei does not modify the MW .S to the extent

, that the validation testing to date would be invalidated, at which point further testing would have to be perfornal by the COL hcider.

SSAR Revision: NONE 820.22a W we*chouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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!1! 'l Question 620.27 The first untence of the firat full paragreph. uxond column of p. P18.6 2 appears to be missing a word. Clarify the wntence (Section 18.6),

Response

The wntence in the $$AR will be nudified as follows:

SSAR Revision:

'Ihe circle in Figure 18.61 represents the dixision wts that are usually mic by the Combined Licene holder relative to plant operations.

l c20.27-1 W westinghouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION Question 020.30 Section 18.6.5 of the SS AR states that '...this model (the User Behavior / Decision Model) of the process muds to be the mental nadel that is poucawd by the operators. Therefore, the mental nxxtel must be reinforced in the interfacing resources arxl taught in the training program...' What is the rnental nxxlel operators of conventional nuclear y>wer pianta pouess? Ilow doca it differ from the propaed mental nxdcl that will be created by the User Behavior / Decision. Making Model? What impact, in terms of learning new behaviors and unleaming old behaviors, will this new nxxlel have on AP600 operators? Ilow difficult a teaming proccan is anticipatal for olerators to develop md become proficient at using the correct nental nwxtel required to effectively operate the AP600 (p. PIB.6 5)?

Response

It is not known what mental nxxlel conventioaal nuclear power plant operators poucas. It protably ranges over quite a large fraction of the pnsible axxlets. At one extreme is a puxici that is bamt miety on the interface that is used each day. This nxxlel is most likely to be a ' pattern recognition' model that ' understands

• the plant as a pattern of colored lights and meter indicator locations. As long as the pettern is ' normal', every thing is alright.

Reasomng stout abnormalities is extremely hmited because there is no buis for the teamming process. This axxlel is likely to be typical of operators that feel that their job is to read the procedure and carry it out, with no emphnis on t,ndenlanding why they do what they do or why the prxedure is what it is. Here is little if any technology transfer between the plant system designers and the plant operators.

In the middle, where nxnt operators are likely to fall, is a mental nxxlel of the physical layout of the plant that enables some types of reasoning. For example, issuca about physical connectedness of plant components (e.g., can I get water from point A to point B7) are well supported, but the nuxlel is hmited with respect to an understanding of the cornequences of actions or for the evahtation of1roblems (e.g., the limitations of a closed valve for energy imbalance and corwequential heating).

The other extreme is operators with mental uxxlels that are ckwer to that of ' snt system designer. lie understands the plant pnxesses and the relationship between the equipment and thr . .ceuca, i.e., what purpcnes the equipment is designed to perform and if not performed what other purposes are being affecta!. lie is able to qualitatively asseas plant state without a procedure and to reawn alcut abnormalities. Procedures are a supplement that remind him of alternatives and the details of system operation. They nmi not be exhatu.tive in their coverage of shmormalities. He ytxxl operalois in current plants, we believe, have a nxxlel that is close to this, but they learned .t or developed it through their own talent and experience m an ad hoc manner. As a result, many operators have ga;" n their mcxlel of this ;ype and when those gaps appear in the course of operations, they must rely on sonening cir.e, such as the interme design, to carry them through. As a result, they are no longer in a position to be mentally ahead of the pnxess they control and cannot effectively evaluate a proceduie for its correctness or effectiveness. Therefwe, if it, too, is flawed, an abnormality in the plant process is the likely result.

c20,20.,

w wesmuse

hnC ftE0 VEST FOR ADDITIONAL INFORMATION Currently, there is no systematic effort to develop and reinforce a functional mental nudet of the plant nere is little esplicit training or tuting of a functiotal understandmg of the plant and the pniceduru and control room interfaces do not support this type of thinking.

The intent of the AP600 M MIS de4ign effort with rupoct to mental nvxlels is two fold. First, the goal is to support reauming about plant proccue4 at a f unctional as well as a physical level. Scumd, the goal is to develop and reinforce a systenutic decision pniccas . hat is bawx! on the Itammussen User Behavior / Decision Making nudel, Specifically, the goal is to insure that operators maintain a broad view of plant proccanes, develop exivstations cf i

process behavior, ched for disconfirming evidence, and to consider aide-effncts of actions, %e deAign objective '

is to develop these mental nydels through training, and hi reinforte them through displays and procedures, One purpose of the User Behavior / Decision Making Modelis, ide411y, to have the plant's system designers transfer their knowledge minut the behavior of the plant through the nudel to the M MIS designers for the dedgn of the interface and then to the operators w> that they willlave, from the beginning of their training a consistent and accurate nxdcl to lecen and one that is reflected in the interface that they use, in this way we hope to narrow the spread of mental nxxtels that exist with the operators, thereby, noving the mean of AP600 operator [wrformance much closer to that of hday's beat operstars.

The questions of learning new behaviors, unlearning old behaviors and the difficulty of the learning pmcess is largely unknow twcause not enough research focus has twwn placed on cognitive skills, it depends upon nany f actors, among them are such things as untivation and the relative closeness of the relf taught /scif4evised nodel that the individual operator currently has to the U6er Behavior / Decision Making Model de4 ired. To nome degree, esperience in this area is leginning to le obtainest with the We6tinghouse up-grade of the computer system and control nom of the Bernau units in Switrerlead.

SSAR Revision: NONE G20,20 2 W Westinghouse

I NRC REOUEST FOR ADDITIONAL.INFORMATION Question C20.31 Section 18.6.5 of the SSAR states that 'the nxdelling pnxess satis 6es the requirements in NUREO-0700, Appendis 11.... This pror. ras is used by Westinghouse in the design of computerized operator support sys'eris...'

What is the pniceas being referred to here (p. P18.6 6)?

Response

The 'pncesa' ticing referred to is the process that builds the User lichavior/Dxision Modelor ' goal means' nxxlel.

This trxdel la equivalent to the one descrited in NUREG-0700, Appendix D. The NRC provides an example of such a goal means axdcl in Eahibit B 2 on page 116 of the appendis.

Westinghouse first uwd a prtcess similar to the one descriled in the AP600 SS AR to develop a goal-means nxxlet of a Pressunted Water Reactor in the ' 'j 1980's in support of the design of the Westingh*mse Safety Parameter Display System. Such a nxdel was originally developed Ny Westinghouse in work performed for EPRI in examining the feasibihty of a plant wide Disturbance Analysis and Surveillance System and is reported in EPRI NP-2240,1982.

SSAR Revisiom NONE 620.3M W Westinghouse t

I NRC RTQUEST TOR AUDITIONAL IFORMATION ill" li!C W ili Question 620.32 Settmn 18.6.7 of the SSAR states that 'the preceding discuuion has examined three unicis that are used for various tek arnlysis.' What are the three rmdels being referred to here? De precedmg discuuion appears to refer to two models: the

• Decisions Sets Organization Model,' and the ' User Behavior / Decision Making Model.' To further mmplicate matters, the term 'model* is also uwd in refernng to Rasmunen'a 'means-end model*

(p. PIB.6 4) and a ' supplementary model" (p. pl8.6-6) as well. Clanfy inis statement (p. P18.6 7).

Response; The nutels being refened to mclude:

1.) The Dnision Sets Model, which looks at the demands placed upon the M MIS from a more global plant perspective. There are addi;ional decisions that the owner of a nuclear power plant must make relative to the operation of the plant that are not included in the real4ime decision Ser. These include, but are not hnuted to, collecting and analynng pnwess data relative to the needs of reliability-centered or predictive maintenance, monitonng and recording personnel radiation esposure, collecting, storing, retneving, and analping process data for engincenng purposes, supporting additional advisors with appropriate data during planc emergencies, supporting spare time tasks of the control room staff, such as up-dating procedures, alann logic, and drawings, administrative and management taska inclading preparation of proecss data for regulatory and safety organizations, etc. Many of these tasks need to be supported by the M-MIS, but the tuks are not identified nor are they defined by the s,econd and third models. it is the objective of this first model, the Decision Sets Model, to establish a framework or Some sort of structure for this aspect of designmg an M MIS.

In the initial phase of the M MIS design prccess, this unlei begms its esistence as little more than a list of known tasks, such as the one in the previous paragraph, that are done by a typical utility that is operating a Westinghouse PWR. His model is continually refined dunng the early steps of the design procca, until the steps involved with prepanng the M MIS System level Functional Requirements are reached. Along the way, this matel has incorporated the results of the efforts of the two remaining matels, as well as any differences related to the overall operation of the AP600. By the time the M MIS design process demands a set of System level Functional Requirements, this first model is refined to the point that it renects the design decisions relative to the scope of tasks supported by the M MIS and represents a reasonable architecture of how the tasks are related. At this point, the M MIS designers can begin to see the data that must be conveyed between decision tasks and the support (e.g., calculational) that the decision tasks need. These inues form the Design Basis of the M MIS and are reflected in the M MIS System Level Functional Requirements so that hardware and software engineers can understand the demands of the M MIS.

m.32a W weSuneme j

l NRC REQUEST FOR ADDITIONAL.lNFORMATION W~M i:

2.) The User Iklunior/Dnision Making blodd or meam-ends modd, works with the third m:xlel in that it provides a compatible nulel of the plant's prmenes. Rasmunen notes (lat expert real time picm.s control operators are rnost effecthe when they have a mental unlet of the proccues they control that is made up of the functions or purposes of equipnwat they control and the links that exi6t between the functions. For example, a valve can 6 tart or stop or nvxlulate flow. He cont.cquences or links are that when the valve position is changed, flow changes. AND a down stream reservoir's level is changed OR mohng rates are changvd wbich, in turn, affcct energy content and, therefore, temperature. Understanding the complete set of links of this type is what the 'means ends' model or the User Behavior / Decision Sets nnlel is all about. Note ti at some cogmtive pinhologists have called these links the ' context

• within which praea data must be preacnted in ordn to msure that the data is properly interpreted and understood. Applymg the results of Remussen's Decision Making model to the mxles of the 'means-ends' model, as noted in Figure 18.613, p. P18.6 20, guides the M-MIS designer in collecting the necessary parameters, setpoints, and synthetic variables about th- ' nt's proccues for supporting the real tinw decir. ion snaking tuka that the plant's prtwesses demand, regardless of whether the decisions are made by automatir control systems or by human operators. Thus, this model and the third model work hand in hand to guide the collectmn and appropnate organization of the real time proccu control data, prior to any M-MIS design decision relative to tuk athwation between man and machine, ne reference to Rasmunen's 'means ends

• model is used in the discuuion to orient a hunan factors reaJer of Rasmunen's work to the fact that Rumussen desenbes the U6er Behavior / Decision Making Model as n *meam-ends' nniel. Within the linuts of the subjectiveneas of the m(xleling proccas, these are identical types of matels.

3.) Rasmuwn's Deri. ion Making Model of Monitoring, Plardng, and Controlling that led Woods and llotinagel to their Stmeture of the Real Time Analysis shown in Figure 18.6-13 (p. P18.6-20) or supphsnentary model. He term ' supplementary r xxirl' is us n P18.6-6 as a way of introducing Rasmussen's Decision Making Model. Rasmussen's nwxlel ' supplements' the other two unlels by adding the human psychological buu to support them, thus providing an understandmg of the cognitive needs of the hunum decision makers with regard to the real-tmw control of the plant's proccues. He Wmxh and llotinarel work uses Rasmussen's moJJ ) determine the questions that a hunan operator must answer in order to monitor plant pnxeues, plan any corrective actions, and to execute any control actions, i.e.,

provide the mechanism for interpreting the other two models so as to form the functional requirements for the design of the data presentatma portion (the control board) of the Ap600 M-MIS.

SSAR Revision: NONE G20.32 2 3 Westingt10Use l

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O NRC REQUEST TOR ADDITIONAL. lNFORMATION Question 620.30 Who performed, or is performing, the verification and validation activitica (for Phnca I and 2) discussed in Section 18.8 of the SSAR7 What art .neir roles (i.e, how did they perform these activities)? What is the proccas used to accompli 6h them (p. P18.8 21, Figure 18.8.2-5)7

Response

he verification and vahdatmn activitica described in the SSAR are being conducted by a multi-disciplinary team that includea: 1) man machine interface design specialists from the Man hiachme Design group at NATD with exteraive experience in designing computerized systems and displays for Nuclear Power Plant control room application; and 2) specialists in human factors, applied psychology, and cogr?'.a engineering from the ilumun Sciences group at the Westinghouse Science and Technology Center. These two groups are primarily responsible for the design and implementation of the hi hilS verification and validation plan. In performing their work they have access to and support from multiple sources of expertise including personnel with expertise in: AP600 system and l&C engineering; AP600 procedures development; AP600 training development; and AP600 human reliability analysis and probabilistic riti asseasment.

The verification and validation plan described in the SSAR was pnmarily developed by staff from the iluman Scieneca group at the Westinghouse Science ani Technology Center, who have extensive experience in analysis and nuicSng of Nuclear Power Plant (NPP) operator perforrrumee in emergencies, human reliability analysis, and the devewment of training programs and decision-aids to support operator cognitive performance.

The process that they used to develop the verification and validation plan is summarized schernatically in SSAR Figure 18.8.2 5 and the accompanying text in SSAR Subsection 18.8.2.3. Briefly a set of rnajor human performance issues that need to be evaluatcd as part of the AP600 verification and validation was identified (Phase 1). For each issue a set of verification and validation tests was then dermed to establish that the AP supports the human performance requirements (Phase 2). In developing the Verification and Validation plan the lluman Sciences group drew heavdy on 1) the Rumussen model of operator decision making as extended by Weatinghouse; 2) their own empirical rescatch and modeling of NPP operator performance in emergencies, the sources of cognitive demands and potential for human ermr; 3) the gene al Iluman Factors literature, including recent results from related industrica on the pitfalls associated with new automation and display technologies (e.g.,

experience in the aircraft industry).

He verification and validation plan generated by the !! unum Sciencea group was then subjected to several review and revision cycles that included critical input from the hiraAtachir.e design groop; the AP600 I&C design group; and the AP600 management team. The studies specifi d in the verification and validation plan will be implemented by a multi disciplinary team that will include staff from the Westinghouse STC Iluman Sciences group who will provide technical expertise in design and analysis of ti e evaluation tests; staff from the NATD hian htachine group c20.38.,

i w wesunsouse

- e NRC REQUEST FOR ADDITIONAL INFORMATION I

who v4JIle responsible for die design of the M. MIS computerized displays, alarna and procedurc4 to le used in the V&V tests; and stalf from the procedurca development, training, and protaldtistic ri6k aucurtwnt groups, who will provide tuhnical guidance in dc4igning test scenarios, prmdures, and training naterials to be uxd in the VAV tests.

SSAR Revision: NONE s20.ac.2 W westinpouse

o NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 620.44 Are the ITAAC provided in Revision 0 of SSAR Chapter 18 cont.idered to be inclusivz/wmplete for M MIS (Section 18.8, p.18.8 26)?

Response

The act of ITAAC provided in SSAR Chapter 18 is inclusive / complete with respect to human performance issues.

SSAR Revision: NONE l

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NRC REQUEST FOR ADDill0NAL INF0ftMATION p.m >g

! ! y Question 620.48 Eanctly how does the alarm system unt the requirements of the SPDS7 Where is the SPDS function located in the control room? Does it nut all the requirements for SPDS as specified in NUREO-0737, Supplement I and amplified in NUREO 1342 ('A Status Report Regarding Industry implementation of Safety Parameter Display Systems,' 1989) (Section 18.9.2.2.6, p. PIB.9-6)7

Response

Our po6ition with regard to the SPDS in the AP600 control room is the same as that reviewed by the NRC for the

$P-90 or Advanced Prenurized Water Reactor (APWR) in the late 1980's. This position, in turn, is a derivative of th SPDS design activity that Westinghouse submitted for NRC review in the early 1980's and which received an SER on the generic or non-plant specific design as cited in the SS AR. Our position is that the need for an SPDS is really an indictment against the performance of the alarm system at nree Mile Island Unit 2 during the March, 1979 incident. That indictment, as stipulated in NUREG 0696, suggested that better performance of the alarm system could have been had during the TMi 2 incident if the alarms had been better organized (around the concept of Critical Safety Functions), had cause-effect relationships been more clearly presented, had fewer alarms been presented (aggregate and abstract detailed alarms to more clearly show the current overall state of the plant's proccuca), and show, in an analog fashion, the plant's proceses deviating from their capected normal states prior to reactor trip.

The Wettinghouse AWARE System, which is the basis for the AP600 darm system does all of these, with one notable eaception. The User Behavior / Decision Making model is the basis for the organization of the alarm messages. This model providea the mapping to the Cntical Safety Functions. The number of alarm messages presented to the operatora is vastly reduced by using a queuing scheme based upon the User Behavior / Decision Making model (i.e., messages are formed to answer the questions als Figure 18.613, p. PIB.6-20, and queued as per the User Behavinr/ Decision Making nodel) and the most urgent are provided the most salient presentation wpace based upon an onC prioritization scheme that is again primarily based upon the

• goal means' (User Behavior / Decision Maiiag) model, ne notable eaception, as was noted in the NRC review of the SP 90, is the fact that an alarm system is not an analog presentation sjstem. It is binary. A menage is presented when a setpoint is reached, off or on. During the review of the SP-90, the NRC noted that the control board CRT displays, also based upon the User Behavior / Decision Making model, did provide the analog presentation of the plant process state and is highly coordinated with the alann presentation 60 that in combination, the alarm system and the display system together achieved the intent of the SPDS requirements. His is the objective of Westinghouse with these new plant designs.

We feel that the requirements of the currently stand-alone back fit devices need to be cammined as part of the total requirements for the design of the M MIS and integrated in a complete sense into the appropriate portions of the M MIS, not left as stand alone devices. Included in this overall integration effc

• for the AP600 are the SPDS (NUREG 0696) requirements, the By-Passed and inoperable Status Indication System (Reg. Guide 1.47), and Post-Accident Monitoring System (Reg. Guide 1.47).

SSAR 9evision: NONE W-WBSilflgh0US0

NRC REQUEST FOR ADDITIONAL INFORMATION

!i!" iII

n Question 620.49 C!stify how the cenclusions about the AP600 alarm system were denved (Section 18.9, p. P18.9 8 through P!s.g.13)7

Response

SSAd Section 18.9.2.4 (p. P18.9-7)is ' Alarm System Design Basis'. The subuquent sub-sections are not as much

' conclusions

• about the alarm system as they are functional Requirernents. %ese Functional Requirements are detailed design objectives for the AP600 alarm system. *f hese Functional Requirem:nts for the alarm system are included here in order to provide an example of the output of the ' Functional Lequirement' step in the M. MIS design process.

The individual Requirements are the result of many things, including, but not limited to: experiments, customer prefererwes (e.g., the ALWR Utility Requirements), principals of applied cegnitive psychology, human factors standards and guidelines, technical society standards (e.g., the Instrument Society of America standard on alarm systems), NRC guidelines and requirements, experience with currently operating alarm systems, and engineering judgenwnt.

SSAR Revision: NONE s20 A s.1 W Westingtiouse

l l

1 NRC REOUEST FOR ADDITIONAL INFORMATION Ouestion 720,49 Provide a snore complete accounting of the source term input to the M ACCS code for each relcue clau, including the relene fraction for each fission product group, and the time, duration, energy, and elevation of relesw.

Response

ne information requested can be found in the AP600 PRA Appendix M. Table M -2, page M - 6.

PRA Revision: NONE 720.49 1 W

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