ET-NRC-92-3789, Forwards Westinghouse Response to NRC 920923,1001,09 & 28 Requests for Addl Info on AP600

Forwards Westinghouse Response to NRC 920923,1001,09 & 28 Requests for Addl Info on AP600

ML20126E240
Person / Time
Site:	05200003
Issue date:	12/22/1992
From:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-NRC-92-3789, NUDOCS 9212290065
Download: ML20126E240 (108)

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ET.NRC 92 3789 NSRA APS192 0270 Docket No. STN 52-003 Prey to'1b December 22,1992 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: DR. T110 MAS MURLEY SUlHECT: WESTINGil0USE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMN110N ON Tile AP600

Dear Dr. Murley:

Enclosed arc three copics of the Westinghouse responses to NRC requests for additional information y on the AP600 from your letters of September 23,1992, October 1,1992, October 9,1992 and =

October 28,1992. This transmittal is a partial ressmse to those iciters. A listing of the NRC requests for udditional information responded to in this letter is contained in Attachment A. The Westinghouse responses to the remainder of the requests for additional information contained in your letters of September 23,1992 and October 1,1992 will be provided prior to January 23,1993.

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

A Nicholas J. Liparulo, Manager Nuclear Safety & Regulatory Activities

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ET NRC 92 3789 ATTACitMENT A AP600 RAI RESPONSES SUBMITTED DECEMBER 22,1992 RAI No. Issue 100.003 i Operational assessment 210.001 l Safety and seismic classification 210.0(M l Reg Guide 1.26 compliance 210.008 i Piping analysis 210.011 l ISM piping analysis 210.015 i Computer codes for stress analyses 220.015 l Containment penetration reinforcement 230.006 l Damped seismic design response 250.009 i Pipe weld CUF 250.011 i Access points 250.012 i Tubesheet handholdes 250.013 l Indexing 250.014 ) U bend area access 250,015 i Remote inspection features 250.017 l Eddy current inspection 250.018 l Leak corrective measures 250.019 l Material strength properties 250.020 l Exceptions to Reg Guide 1.121 251.032 l SRP compliance 252.003 l Analyses 252.004 l Piping material property verification 252.005 l Class 2 & 3 piping 252.006 i Piping outside contairiment

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252.010 1 Piping stresses for different sites 252.011 l RCS piping stresses 252.047 l ASS carbon content analysis 1

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1rr NRC-92-3789 ATTACllMENT A AP600 RAI RESPONSES SullMrlTED DECEMBER 22,1992 RAINo. Issue 252.060 l Inaccessibility to cavities / chambers 252.107 i Tube plugging criteria 252.108 I Facilitating SG tube fusion techniques 252.109 l Tube vibration wear 252.I11 l Misplaced anti vibration bars 252.112 l Delta 75 SG primary side manway size 252.113 l Records / material archive program 252.114 l AVT secondary water chemistry with Inconel 690 252.115 l Stress corrosion cracking in Inconel 690 252.116 l Inconel 690 resistance to corrosion 252.137 i Materials 252.143 l System failure jeorgrdizing safe plant shutdown 420.005 l Annunciators, manual actions & defense-in depth 420.006 l ICS common rnode failures 435.005 i Post 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portable diesels 435.007 l Reg Guide 1.75 separation of nonsafety ac division 435.009 i Diesel generator maintenance / testing program 435.012 l Lead sizing 435.017 l Lightning protection of main setup transformers >

435.024 l Overvoltage protection of batteries 435.025 I Ground detection of Class IE de system 435,032 l Reg Guide 1.75 compliance 435.033 l de power system transient respons:

435.065 l Circuit breakers used for electrical isolation 435.068 l lE/nonlE boundary ac power source /120 Vac bus 435.069 l Emergency lighting requirements 2-

LT-NRC 92-3789 NITACHMENT A AP600 RAI RESPONSES SUllMITTED DECEMBER 22,1992 RAINo. Issue-440.022 l Fuel design and analysis 440.023 l Mid-loop operations 440.030 l Interfacing system LOCA 450.009 l Sump pil control 471.002 l Radiation zone designations 471.003 l ALARA concerns 480.001 l Operator actions during 72-hours post-accident 480.003 i Failure modes for PCS 720,003 l Failure modes 720.006 l PRA methodology 720.007 l Truncation limits 720.010 i Human error probabilities 720.016 i PRA assumptions 720.023 i PRA assumption concerning PCS availability 720.026 l Containment drain design 720.030 l MAAP 4.0 720.035 i Hot leg creep rupture 720.036 l Hot leg creep rupture 720.038 i Fuel-coolant interactions 720.039 l Core-concrete interactions 720.045 l Fission product holdup 720.046 1 Fission product transport and retention 720.047 l- Xenon & Krypton release fractions 720.050 l Sensitivity & uncertainty for MAAP analysis 720.057 i PRA-720.058 i Data files 3

7 J NRC REQUEST FOR ADDITIONAL INFORMATION Question'100.3 An applicant for a standard design certification is required by 10 CFR Part 52 to address issues identified as a result of the TMI 2 event (10 CFR 50.34(f)), unresolved safety issues (USis) and generic safety issues (GSis) that have been prioritized an either high or med;um, as well as ell applicable rules and regulations.~ The NRC staff has. as?

part of its overall design certification review process, been reviewing operational experience information contained in NRC bulletins and generic letters issued since January 1980' to identify any potential safety issues that should be considered during the NRC staff's design certification review but ray not be specifically addreased thru any of the mechanisms highliahted above.

De staff has reviewed the SSAR to determine'whether operational experience informathm has been effectivily --

incorporated into the AP600. De staff has concluded that the discussion provided in Section 1.9 of the SSAR does -

not contain sufficient detail for the staff to detennine that operational experience information has been effectively incorporated into the AP600.

Provide additional details regarding incorporation of operational experience information imo the AP600 design.

Specifically, incorporate a tabulation of NRC bulletins and generic letters issued since January 1980 into Section 1,9 -

of the SSAR. This tabulation r.hould include the applicability of each bulletin and generic letter to the AP600, as -_

well as its disposition. Typical dispositions could include a) a determination that the issue identified in bullets or -

generic letter is not applicable to the AP600 and the basis for that determination, b) a determination that the issue identified in the bulletin or generic letter will be addressed as part of a USI, GSI, Rule, Regulatory Guide, or TMI -

Action item,' including references to Imth the applicable USI, GSI, Rule, Regulatory Guide, or TMt Action item .

and the appropriate section of the SS AR where the issue is or will be addressed, or c) a determination that the issue -

identified in the bulletin or generic letter :is or will be covered by interface requirements for_ the' COL applicant.

In addition, a description of any enhancements made to the AP600 based on other sources of oper_stional experience information (with' appropriate references to the SSAR) should also be provided.

Response

This RAI has been addressed by; WCAP-13559, ' Operational Assessment for AP600", December,1992. Submitted December 15,1992 in letter ET-

. NRC-92 3784.

~SSAR Revision: NONE 1NUREG-4690, Vol.1, .Rev.1 " Generic Communications Indc6 provides a listing of NRC 4 generic'-

communications (bulletins, generic letters, circulars, and information noO es) issued from 1971 to 1989.-

' 100.3-11

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 210.1 Discuss thejustification for the safety and seismic classification of structures, systems, and components (SSCs) that are unique to the passive design of the AP600 (i.e. Passive Core Cooling System, Passive Containment Cooling System, etc.). The staff is concerned that there is no previous experience with these systems and they fall outside the structures, systems, and components that have traditionally been classified utilizing the guidelines set forth in Regulatory Guides 1.26, and 1.29 for safety and seismic classification, respectively (Section 3.2).

Response

The safety philosophy of the AP600 is basically the same as current PWRs, with the safety-related systems performing the safety-related functions. What is different is that the safety-related functions are performed by different systems that are made up of different arrangements of piping, valves and components. Regulatory Guides 1.26 and 1.29 are functionally defined for the most part and therefore they have been applied to the safety-related AP600 systems. There are two areas of the AP600 design where these Regulatory Guides do not allow a straight-forward classification, as discussed below.

Subsection 3.2.2 of the SSAR describes the AP600 classification system. This system is very similar to and is compatible with Regulatory Guide 1.26. Most of the AP600 structures, systems and components have the same Regulatory Guide 1.26 classification. For example, the AP600 RCS components and piping are class A, which is equivalent to Regulatory Guide 1.26 quality group A. The AP600 containment is class B, which is equivalent to Regulatory Guide quality group B. The AP600 classification system differs from Regulatory Guide 1.26 in the foilowing two aspects:

1) Portions of safety related systems that interface with the RCS or containment are considered to be class C if they are not part of the RCS pressure boundary and they do not recirculate fluids from the RCS /

containment outside of the containment. Regulatory Guide 1.26 requires quality group B for an ECCS system, but only requites quality group C for an auxiliary feedwater system.

The AP600 classification requires a portion of a safety-related system that is part of the RCS pressure boundary to be class A, as in the current Regulatory Guide 1.26. If it recirculates post-accident fluid outside of the containment, it would be class B, as in the current Regulatory Guide 1.26. None of the AP600 safety-related systems that interface with the RCS or the containmeut recirculate fluid outside containment. Several portions of the PXS systems are class A because they are part of the RCS pressure boundary, including the PRHR llXs and the CMTs. The IRWST and the accumulators are class C because they are ECCS camponents located inside containment which can not recirculate post accident fluid outside of the containment. The passive containment cooling system water storage tank and its associated piping and valves are class C because they provide a safety related function and they are located completely outside of the containment. This system only contains nonradioactive water.

2) Nonsafety-related systems that mitigate events are not addressed in Regulatory Guide 1.26. The AP600 classification system addresses this type of equipment, it uses a special nonsafety-related classification, l

W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION-0

- Class D, which applies to both defense-in-depth equipment that reduces the potential for passive safety -

related system actuation and to radioactive waste processing equipment. - This class is equivalent to.

Regulatory Guide 1.26 quality group D. For example, the startup feedwater system is clean D because it has no safety-related functions, but does provide decay heat removal for events such as transients and loss of outside power.

SSAR Table 3.21 provides a comparison of the AP600 classification system with Regulatory Guide 1,26, as well-as ANSI 51.1, SSAR Table 3.2-3 provides a listing of the classification of the equipment'and valves in the AP600 :

systems. Appendix 1 A of the AP600 SSAR discusses compliance with Regulatory Guides.

. :i SSAR Revision: NONE +

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 210.4 The last sentense of the first paragraph in Section 3.2.2.1 of the SSAR states that "These definitions are consistent with the draft ANS Definitions for LWR Standards.' These definitions should be explicitly identified in the SSAR since the staff does not presently endorse an ANS standard for the classifications of structures, systems, and components. He staff relica on Regulatory Guide 1.26 for that purpose. Provide technicaljustification for any deviations from Regulatory Guide 1.26.

Response

The definition of the AP600 classifications is provided in SSAR section 3.2.2. These definitions are similar to Regulatory Guide 1.26. Refer to the response to RAI 210.1 for a discussion of the differences. Appendix I A of the AP600 SSAR discusses conformance with regulatory guides.

SSAR Revision: NONE Yj Yl85tiflgh0US0

NRC REQUEST FOR ADDITIONAL.INFORMATION Question 210.8 Section 3.7.3.8.2.2 of the SSAR states that for ASME Class I piping equal to or less than one inch nominal pipe size and ASME Class 2 and 3 piping equal to or less than two inch nominal pipe size, one of the following three methods of analysis may be used:

a. 7he method for large diameter pipe descrital in Section 3.7.3.8.2.1 of the SSAR.

b. Equivalent static analysis.

c. Seismic qualification by experience based on the guidelines in EPRI Report NP-6628, ' Procedure for Seismic Evaluation and Design of Small Bore Piping.'

l l If the procedure for use of the equivalent static analysis as noted in item b above is different from that descriled in Section 3.7.3.5 of the SSAR, revise Section 3.7.3.8.2.2 to provide a detailed description of the methodology to be used.

( The staff is currently reviewing EPRI NP-6628 as a topicalicport, which was submitted to the staff by the Nielcar Management and Resources Council in a letter dated March 19, 1991. Pending completion of this review, the staff's position is that the methodology in this report is not acceptable. Revise Section 3.7.3.8.2.2 to remove the reference to EPRI NP4628.

Response

l There are no differences between the equivalent static analyses described in Subsections 3.7.3.5 and 3.7.3.8.2.2.

We believe the methodology presented in EPRI NP 6628 will be found acceptable by the NRC and should le included in the AP600 review and approval process.

SSAR Revision: NONE 2, o.8.,

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' NRC REQUEST FOR ADDITIONALINFORMATION l

Ouestion 210.11 Clarify the discussion in Section 3.7.3.9 of the SSAR on the use of the ladependent support motion (ISM) nwthod of modal analysis of piping systena to address the following concerns:

a. The proposed ISM method is inconeNtent with the recommendations in Sections 2.3 and 2.4 of NUREO+

1061, ? Report of the USNRC Piping and Review Committee,' Volume 4. Provide further technical justification. As a part of these recommendations, a suppod group is defined by supports that have the same time history input. This usually means all supports located on the same floor (or portions of a lloor);

of a structure.

b. The damping values in Section 3.7.1 of the SSAR are referenced for use with the ISM method. ' His implies that the AP600 design incorporates the use of ASME Code Case (CC) N 411, " Alternate Damping T Values for Responso Spectra Analysis of Classes 1, 2, and Piping, Section !!!, Division l' in corgjunction -

with the ISM method.- One of the conditions in RO 1.84, " Design and Fabrication Code Casei Acceptability, ASMB Section 111, Division 1," relative to the use of CC N 411 is that the stafPs acceptance of the use of the damping values in CC N-411 with the ISM nethod is pending furtherjustification. Since

' the proposed ISM method is not in'accordance with the recommendations in Item a above, provide further' technicaljustification for this approach.

~

Responso:

a. Technical justification for the proposed independent support motion (ISM) method is a comparison with test results as reported in EPRI NP-6153, " Seismic Analysis of Multiply Supported Piping Systenu',. j-Project 96410, March,1989. A support group is defined by supports that have the same time history :l input. This usually means all supports located on the same floor (or portions of a floor) of a structure..

b. ASME Code Case N 411 damping values will not be used with the ISM method. This satisfies the requirements of Regualatory Guide 1.84. Conformance to Regulatory Guides is addressed in SSAR Section -

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.j 1.9. 'Ihe damping values for piping systems that are analysed with the ISM method are 3 % for piping'-

larger than 12 inch diameter and 2% for smaller piping.

The following changes will be made to Subsection 3.7.3.9 of the SSAR:

SSAR Revision:

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Lupport' group is defined by supports'that have 'ths name time :

history input; This usually means'all supports located on the same floor (or portions of a floor) of a structure.The ,j SSB ' damping values'for piping systems that are analysed with the ISM method.are 3 % for piping larger ,than_1.2 - 1 inch diameter and 2% for smaller piping, y

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NRC REQUEST FOR ADDITIONAL INFORMATION WE Ouestion 210.15 The guidelines of Paragraph II.2 in Section 3.9.1 of the SRP state that a list of computer prognuns that will be used in dynamic and static analyses to determine the structural and functional integrity of seismic Category 1 Code and non-Code iterns, and the analyses to detennine stresses should be provided. Provide such a list. Also, discuss the various programs' applicability and validity. At present. Section 3.9.1.2 of the SSAR only references the quality assurance prognun (as described in Chapter 17 of the SS AR) for this information.

Response

The list of computer prognuns that are used for seismic Category I toechanical components is provided in Table 3.9-15 along with the application of each program. Other prognuns inay also be used to complete the dynamic, static or stress analysis of these components. The computer programs will be listed in the ASME Code Design Report.

The validation of each prognun is in acconlance with an established quality assunmee program and is av:ulable for NRC audit.

Subsection 3.9.1.2 of the SS AR will be resised and Table 3.%15 added as follows:

SSAR Revision:

A number of computer prognuns that are used in the dyn;unic and static analyses of mechanical loads, stresses, and deformations of seismic Category I components and supports are listed in Table 3.9-15. A complete list of prognuns is included in the ASME Code Design Report.

The development process, verification validation, configuration control and error reporting and resolution f or compuur prognuns used in these analyses for the AP6Whtre completed in compliance with an established quality assurance prognun. The quality assurance prognun is described in Chapter 17.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION-Table 3.9-15 3

Computer Programs for Selsmic Category l Components Program Application A11AQUS Finite element structural analysis ANSYS Finite element structural analysis CAEPIPE Static analysis of piping systems -

FATCON ASME fatigue analysis of piping components GAPPIPE Static nnd dynamic analysis of piping systems MAXTRAN Transient stress evaluation of piping components PIPSAN Structumi and ASME stress analysis of component supports PS+CAEPIPE Static and dynamic analysis of piping systems STAAD-Ill Static and dynamic analysis of stmetural frmnes TilERST Transient heat transfer analysis of piping com[xments WECAN Finite element structural analysis WEGAP Dynamic structural response of the teactor core WECEVAL ASME stress evaluation of mechanical components WESTDYN Static and dynmnic analysis or piping systems 210.15-2 3 Westinghouse -

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NRC REQUEST FOR ADDITIONALINFORMATION k' h Question 220.15 Det.cribe how penetrations and penetration reinforcements will be analymt for buckling. He area-replacement rule may natisfy tensile strength requirements, but it does not necenarily satisfy buckling requirenwnts (Section 3.8.2.4.2.5).

Responso:

Re penetrations and penetration reinforcements have been designed in accordance with the rules of ASME 111, Subsection NE. References 220.15-1 thru 220.15 3 provide data which shows that penetrations with 80% or more of the area-reinforcement required for the internal pressure have the same buckling strength as an unpenetrated cylinder. Therefore, based on extensis e experience with operating units and the results ofin-house testing prograna, no further buckling anessments of the penetrations and penetration reinforcements are deenni appropriate.

REFERENCES 220.15-1: C.D. Miller, " Experimental Study of the fluckling of Cylindrical Shells with Reinforced Openings," ASME/ANS Joint Conference, Portland, Oregon, July 26-28,1982.

220.15-2: C D. Miller, R.B. Grove, " Buckling of Cylindrical Shells with Reinforced Circular Openings under Axial Compression,' Internal Report by Chicago Bridge & 1ron Company, Plainfield, Illinois, March 14,1980 (Letter Liparuk to Murley, ET-NRC-92-3778, December 2,1992).

220.15-3: J.G. Hennett, R.C. Dove, T. A. Butler, "An laveatigation of Buckling on Steel Cylinders in Circular Cutouts Reinforced in accordance with ASME Rules," Los Alamos Scientific Laboratory, LA-8853 MS, NUREG/CR-2165. June 1981.

SSAR Revision: NONE W Westinghouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION illi! !ii :

!Y' ii Ouestion 230.6 Section 3.7.1.2 of the SSAR states that the *TAFT' earthquake time history was umi to generate synthetic time histories for AP600 seismic design. The SSAR presents spectrum comparison between the Arow damped r.eismic design response spectra and the corresponding RO 1.60 response spectra anchored to 0.3 g for the damping ratios of 2, 3, 4, and 7 % in Figures 3.7.14 through 3.7.1-8. Ilowever, the SSAR should also provide a spectrum comparison for the case with a damping ratio of 5 %. Provide such a spectrum.

Response

SSAR Figures 3.7.14 through 3.7.1-8 are revised to include the 5 % damping response spectrum curves.

SSAR Revision:

(SSAR Figures 3.7.1-6 through 3.7.18 will be revised as shown in attached sheets.)

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 250.9 Table IWB 2500-1 in Section XI of the ASME Code requires the examination of Class 1 piping welds, with a Section 111 fatigue cumulative usage factor (CUF) exceeding 0.4, at every inspection interval. Confirm that the value of CUF to le used will correspond to the projected 60-year plant design life (Section 5.2.4).

Response

no Section lit cumulative usage facto; is calculated on the basis of a 60-year plant design life.

SSAR Revision: NONE W-W85tlagh0llSe

NRC REQUEST FOR ADDITIONAL INFORMAllON Ouestion 250.11 Figure 5.4-2 in the SSAR does not show the orientation and location of all of the access points in the steam generator. Provide drawinga to show the secondary side access points in the steam generator.

Response

There are sia 6 inch handholes located in the lower part secondary ahell. There are four 90' apart (two on the _

tubelane and two at 90' from the tube lane) that provide access to the secondary face of the tubesheet. *lheother two are IH0' apart on the tubelane and provide accean to the top of the flmv distribution balfle. Two 4 inch posts located on the secondary shell in line with the tubclane and above the top tube support plate provide access to the U ilend area. Please see the answer to RAI 250.12.

Figure 5.4 2 will be updated to show the location. The last paragraph of SSAR Subsection 5.4.2.4.2 will be revised to reflect this response as follows:

SSAR Revision:

Several methods can be used to clean operating steam generators of secondary side deposits. Sludge lancing is a procedure in which a hydraulicjet inserted through an access opening (handhole) h>osens deposits and the l<ne material is flushed out of the steam generator. Six 6 inch access ports are provided for sludge lancing. Four of thcae are located above the tubesheet 90* apart (two on the tubclane and two at 90' from the tube lane) to provide (

acces to the secondary face of the tubesheet, and two are 180' apart on the tubelanc and provida access to the top of al-the flow distribution halfle. Also two 4 inch ports located on the secondary shell in line with the tubelane and above the top tube suppcut plate provide access to the U Bend area. A blowdown pipe is provided to permit continuous blowdown and monitoring of secondary water chemistry. The blowdown piping suction is adjacent to the tubesheet and in a region of ielatively low-flow velocity. This facilitates the removal of particulate impurities to reduce the accumulation on the tubesheet. The materials of the secondary side of the steam generator are also compatible with chemical cleaning.

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NRC FIEQUEST FOR ADDITIONAL INFORMATION Question 250.12 Discuss whether the four 15 cm (6 in) handholes located just alove the tubesheet are of suf6cient site to allow for ef fective stuJge lancing, retrieval of hiose parts, and/or inspection of the tube bundle by por'%ble inspution equipment (e.g. video equipment) (Sation 5.4.2).

Response

Two of the four handholen just above the tubesheet are the sarne size and in the same relative location as the handhelen in the Wtwtinghouse designed Mmlel F steam generators. Two additional handholes are providal at approxinuitely 90' froin the ends of the tubelane. The experience in operating plants has demonstrated that these handholes permit ef fective sludge lancing, retrieval of loose objects, and inspection of the tube bundle. Two 4 inch p>rts h>cated on the secondary shell in line with the tubelane and above the top tube support plate provide acces.s to the U ilend area. Please see the ressmse to RAI 250.11.

The last paragraph of SSAR Subsection 5.4.2.4.2 will be revised to reflect this ressmse as follows:

SSAR Rovision Several methals can be used to clean operating steam generators of secondary-side deposits. Sludge lancing is a procedure in which a hydraulic jet inserud through an access opening (handhole) h>osens deposits and the loose material is ilushed out of the steam generator. Six 6 inch access ports are provided for sludge lancing, inspection of the tube bundle by portsble inspection equipment, and retrieval of loose objects. Four of these are located above the tubesheet, and two are above the flow distribution haffic. A blowdown pipe is provided to permit continuous blowdown and nonitoring of secondary water chemistry. The blowdown piping suction is adjacent to the tubesheet and in a region of relatively low-flow vehwity. This facilitates the removal of particulate impurities to reduce the accumulation on the tubesheet. The materials of the secondary side of the steam generator are also compatible with chemical cleaning.

W WestinEhouse 1

NRC REQUEST FOR ADDITIONAL INFORMATION 7,

Question 250.13 Describe the design provisions for tube indexing for facilitation of tube identification and location during innervice inspections (Section 5.4.2).

Response

Controls for the automated, robotic equipment typically used for tule inspection and repair activities deliver the inspection and service tooling to the proper tube location without the need for any visual numbering or other identificatha of the tubes. !!owever, to facilitate tube identification for manual activities, if needed, the tube location for a large fraction of the tubes is scribed on the tubesheet adjacent to the tube. The scribing is done using laser scribing or other method chosen to minimize residual stress in the tubesheet cladding.

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

SSAR Revision:

The steam generators are designed to permit access to tubes for inspection and/or repair or plugging, if necessary, per the guidelines described in Regulatory Guide 1.83. The AP600 steam generator includes features to enhance robotics inspection of steam generator tubes without manned entry of the channel head. Hese include a cylindrical section of the channel head and remote installation of nozzle dams. To facilitate tube identification for manual activities, the tube location for a large fraction of the tubes is scribed on the tubesheet.

250.13 1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION E

Ouestion 250.14 Describe the physical location of the internal deck plates used to gain access to the U-bend area. Clarify the statement in Section 5.4.2.5 of the SSAR that "for proper functioning of the steam generator, r.ome of the deck-plate openings are covered with welded but removable, hatch plates.'

Response

The deck plate through which access is required to reach the steam generator tube U-bends is the deck plate located at the base of the primary separators at the top of the tube bundle wrapper. The deck plate openings in this deck plate may be opened by grinding off the welds retaining the hatch plates in place. Routine inspection requiring access through these hatches is not expected.

The second paragraph of SSAR Subsection 5.4.2.6 will be revised to reflect this response as follows:

SSAR Revision:

The design includes a mimber of openings to proviJe access to both the prinary and secondary sides of the steam generator. The openings include four 18-inch diameter manways, one for access to each chamber of the reactor coolant channel head and two in the steam drum for inspection and maintenance of the upper shellinternals.

In addition, six 6-inch diameter handholes in the shell, four located just above the tubesheet secondary surface, and two located just above the flow distribution baflie, are provided. Two 4-inch diameter inspection opea!.nga are provided at each end of the tubelane between the upper tube support plate and the row I tubes, Additional access to the tube bundle U bend is provided through m4s4-the internal deck platec at the bottom of the primary separators. For proper functioning of the steam generator, some of the deck-plate openings are covered with hatch plates welded inplace that are rM-removable by grinding, gr uging, or other methods to cut off the welds.

[ Westiligh0USB

L NRC REQUEST FOR ADDITIONAL INFORMATION Question 250.15 Dexribe the features incorporated in the design that enhance inspection of the steam generator tubes without manned

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

tube inspection and repair, in addition, discuss whether verification have been performed, by computer simulation and/or mockup, to ensure that the design will facilitate not only the use of roboti6 manipulators in inspecting all of the tubes within the steam generator but also in inserting the robotics into the steam generator (Section 5.4.2).

Response

The cylindrical portion of the channel head just below the tubenheet facilitates the use of robotically delivered -

inspection and repair tooling to tube locations on the periphery of the tube bundle. He ability to reach alllocations has been verified using computer simulations. The channel head and primary inlet and. outlet nor21cs have provisions to facilitate the robotic installation of nozzle dams. - The use of 18 inch diameter manway openings :

provides that any equipment that is used in operating steam generators can be used in the AP600 steam generator._-

- The fourth paragraph of SSAR Subsection 5A.2.5 will be revised to reflect this respome as follows:

SSAR Revision:

-a' The steam generators == b 6:2 te permit access to tubes for inspection; andAw-repair, or plugging, if.

necessary, per the guidelines described in Regulatory Guide 1.83. He AP600 steam generator includes features H to enhance robotics inspection of steam generator tubes without manned entry of the channel head.1%cse include 1 a cylindrical section of the channel head,18 inch diameter primary manways, and provisions to_ facilitate the' remote

~

installation of nor21e dams. Computer simulation 'using designs of existing robotically delivered inspection' and -

maintenance equipment verifies that tubes can be accessed; T westinghouse 2so.1 s-1

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o

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 250.17 Provide clarification on what it considers "more capable equipment' or a " suitable eddy current inspection system

• as compared with the equipment described in paragraph C.2.c of Regulatory Guide 1.83 (Section 5.4.2).

Response

Eddy current inspection equipment currently in us.e typically includes multi-channel capability, software to screen indications, software to provide graphical representations of associated tube degradation and recording on nulia readable on personal computers or workstations. Strip charts and magnetic tape data recorders are typically not used and optical disk storage media may be used. As new equipment and methods are developed for inservice eddy current testing this equipment and testing may be used for baseline inspections. Any eddy current inspection performed in the manufacturing facility is conducted by personnel que.lified to the requirements for inspectors performing inservice inspection of operating units. The manufacturing facility inspection is conducted using the same equipment as, or equipment similar to, that used during inservice inspection of operating units.

(See the response to RAI 250.16 for changes to the exceptions to Regulatory Guide 1.83 found in SSAR Appendix 1 A.)

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

SSAR Revision:

Regulatory Guide 1.83 provides recommendations on the inspection of tubes. The recommendations cover inspec-tion equipment, baseline inspections, tube selection, sampling and frequency of inspection, methods of recording, and required actions based on fmdings. Any eddy current inspection performed in the manufacturing facility is conducted by personnel qualified to the requirements for inspectors performing inservice inspection of operating units. The manufacturing facility inspection is conducted using the same equipment as, or equipment similar to, that used during ' inservice inspection of operating units. Exceptions to Regulatory Guide 1.83 are noted in Subsection 1.9.1.

2so. m W westinghouse 1

1

NRC REQUEST FOR ADDITIONAL. INFORMATION Question 250.18 Describe the corrective measures that will be implemented to disposition leaking tubes, defective tubes, and tubes with imperfections exceeding the plugging limits (Section 5.4.2).

Response

The detennination of the corrective measures to be used to disposition defective or degraded tubes is the ,

responsibility of the combined license holder at the time such degradation or defects are observed. Tubes with eddy current indications in excess of the repair limit may be removed from service by the installation of mechanical tube plugs, by the installation of welded plugs meeting the requirements of the ASME Code,Section XI, IWB-4200, by the installation of tube sleeves meeting the requirements of ASME Code,Section XI, IWD-4300 or other tube repair methods authorized by the NRC. The AP600 steam generator accommodates current and anticipated repair methods and techniques. The tube repair criteria establishing the level at which degraded tube must be plugged or repaired is to be provided by the combined license applicant considering NRC requirements and industry recommendations.

Please also see the response to RAI 250.19.

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

SSAR Revision:

The steam generators are designed to permit access to tubes for inspection and/or repair or plugging,if necessary, per the guidelines described in Regulatory Guide 1.83. Tooling to install mechanical and welded plugs, tube repair sleeves, or effect other repair processes remotely can be delivered robotically. The AP600 steam generator includes features to enhance robotics inspection of steam generator tubes without manned entry of the channel head. These include a cylindrical section of the channel head and remote installation of nozzle dams.

l l

l 250A84 W Westinghouse

i

- NRC REQUEST FOR ADDITIONAL INFORMATION l Ouestion 250.191 Provide clarification to exceptions to criteria C.2.a.(2) and C.2.a.(4) of Regulatory Guide 1.121. - In particular, describe how the proposed change will affect the margin of safety currently observed. Describe the statistical analysis of the tensile test data that is used in the development of the expected material _ strength properties. Also, discuss whether the calculation of the tube minimum wall requirements will be based on the lowcat valuca for the -

material properties, i.e., the lowest values from statistical analyses or from the ASME Code (Section 5.4.2)._

Response

The exceptions noted in Appendix 1 A for Powdons C.2.a.(2) and C.2.a.(4) arr asistent with the methods used '

to develop tube plugging criteria for currently ogwsting plants. Please note the Regulatory _ Guide 1.121 is used to address degradation of steam generator tubes in un that have entered service and has typically not been used - _

to establish the basis for tube plugging criteria prior to operation. The tube repair criteria establishing the level at_

which a degraded tube must be plugged or repaired is to be povided by the combined license applicant considering NRC requirements and industry recommendations. Since the use of the recommendations of Regulatory Guide 1.121 to establish tube repair criteria during the Design Certific< tion process is inappropriate, reference to the  :

Regulatory Guide should be Jrh1

'lhe last paragraph of SSAR Subsection 5.4.2.5 will be revised to reflect isJ' response as follows:

SSAR Revision:

The minimum requirements' for in-service inspection of steam generators, incleding t' beu repair ~'p4*gging criteria, is.the: responsibility off-.the combined ll_icense applicant considering NRC requirements and industry recommendations.5 = :"'": ' = 72:: efi:T 7M :!? :" n'in=. "-- - ':G -- " = r ^ ' ' cit te ^.SME Cch, S:M=4:!, =d ":g !:'n y C i ' !al.Section XI of the ASME Code provides general -

acceptance criteria for indications of tube degradation in the steam generator.- ?; "Ec cc^--nr ii"_g'^' ry.

C !i ' J 2! *: 4ier. -' ' FS- ! 9 Y

-250.19-1 W

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NRC REQUEST FOR ADDITIONAL INFORMATION l f Question 250.20 Provide technicaljustifications for exceptions to criteria C.2.a.(5)-(6) and C.2.b of Regulatory Guide 1,121 (Section 5.4.2).

Response

The AP600 position provided in Appendix 1 A for Regulatory Guide 1,121 Regulatory Positions C.2.a.(5)-(6) s,hould be ' Conforms *. The AP600 position for C.2.b. is that given for C.2.a.(5)-(b).

SSAR Appendix 1 A md the last paragraph of SSAR Subsection 5.4.2.5 will be revised as follows:

(See the re.sponse to RAI 250.19 for related SSAR changes.)

SSAR Revision:

(Appendix 1 A)

Reg. Guide 1.121, Rev. O,8/76 - Baus for Plugging Degraded PWR Steam Generator Tubes C.2.a.(5)-(6) liaception I- 'we-cuffident-impeetion4ata-enic6-ta*4*l44*h Conforms elogadark,nel!c ,x=, :h+,*4e' ' - vwage.6:- ^ d mined feoewthe n.r. ef the k" d^ "b:: ::quirmnenwfor4niainmenwell aw-narkedly-ddferent f*ew+iousereewf-4)*4ubebund!#rce A ::the U-lwnd-*:= == :2reight4engthan-Weatingiu :.: 6 dgn",  ; ::e plagging-limitarM+0 A J '- "' '- - ^ rying-reiquirenwnwin a-nwnnee-whia da c :+64. quire-unnvee ary-plugging 44-4ulwa, C.2..b. . Exception . In cases where sulficient inspection data exist to establish degradation allowance, the rate used is an average tinerate deternined from the mean of the test data. Where requirements for minimum wall are markedly different for various areas of the tube bundle, such as the U-bend area versus straight length in Westinghouse designs, separate plugging limits are established to address the varying requirements in a manner which does not require unnecessary plugging of tubes.

(Subsection 5.4.2.5)

The minimum requirements for in-service inspection of steam generators, including tube plugging criteria, are established as part of the Technical Specifications. These requirements are consistent with the ASME Cale,Section XI,-+nd-Regal *4ory-Guide 4r134.Section XI of the ASME Code provides general accept.nce criteria for indications of tube degradation in the steam generator. Spe4fe-confec==n= cith Regulate:y Cu!& ! !24-ia db__d!,S=9 in 2so.20-,

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__.._..___-.________________,m'

l NRC REQUEST FOR ADDITIONAL INFORMATION It I

i Question 251.32 Discuss conformance with guidance in Acceptance Criteria II.4.a. II.4.b II.4.c II.4.d, and 11.4.e in Section 10.2.3 of the SRP (Section 10.2.3).

Response

The referenced acceptance criteria are stated to apply to built-up turbine rotors. The rotors that would be supplied ,

are fully integral (FI) and the following responses are applicable to F1 rotors.

a. The design overspeed of the rotors is 120 percent of normal speed. The maximum anticipated overspeed, in the event of a trip from full load is 11I pervent.

b. The comlined stress is less than 50 percent of the bore yield strength (See response to RAI 251.27).

c. Turbine shaft bearings are sind to limit the normal gravity load pressures and load additions resulting from misalignments due to thermal effects in the founaation and vacuum loading to conservative levels based on actual bearing tests. We turbine shaft bearings do not experience any abnormal loada due to the transients or accidents as discussed in the RAI questions.

d. The natural critical frequencies are controlled to be removed from normal running speed by approximately 15 percent. He system i.5 evaluated for sharpness of resonance response if separation margin not attained.

Adequate model damping vsures controllable response amplitudes through proper balancing procedures.

e. The F1 rotor does not have any keyways. The rotor may be examined from the bore by existing ultrasonic technique techniques. This will enable full volume coverage of all highly stressed zones.

SSAR Revision: NONE 251224 3 Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION

!h I

.f' Ouestion 252.3 Perform bounding LBB analyses for each of the LBB candidate piping, including evaluations for susceptibility to potential degradation nwchanisms for the projected 60-year plant design life. Provide the analysea (Section 3.6.3).

Response

Bounding leak-before-break analysis is not performed. Sample leak-before-break analyses are provided in Appendix 3B for the reactor coolant loop piping. The NRC stalf should be able to assess the acceptability of the AP600 leak-before-break appmach based on this sample and the criteria in Subsection 3.6.3. The following additional sample calculations will be performed by December,1993.

Pipe Stress Analysie Sample Problems for Piping Evaluated to Leak-Before-Break P&lD number Description Diameter (in)

RCS-M6@l Primary coolant loops (1 and 2) 22 and 31 RCS-M6-002 Automatic Depressurization stage 1,2, and 3 4, 8, and 14 (A and B)

SGS-M6-001 Main steam (I and 2) 32 SGS-M6-001 Main feedwater (1 and 2) 16 The entire scope of candidate leak-before-break pip;ng lines is provided in the response to RAI 210.6.

SSAR Revision: NONE Westingh0US8

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 252.4 Describe the proculures to tw used by the COL applicant to verify that the actual material properties and final, as-built piping analyses are within the limits in the bounding LBB analyses (Section 3.6.3).

Response

Bounding leak-before-break analysis is not performed. He leak-belnre break analysis is performed prior to construction and is based on the as designed piping analyses and representative material properties. A report will be prepared by the COL applicant to reconcile the design analysis. This report will include a review of the as-built piping analyses and the Certified Material Test Reports that are associated with those piping systems that are qualified to the leak-before-break criteria. He pmcedure to develop this report is beyond the scope of design certification.

SSAR Revision: NONE 2s24 1 W westinghouse

NRC REOVEST FOR ADDITIONALINFORMAllot a.

Question 252.5 Sectum 3.6.3 of the SSAR indicaten that Clai,s 2 and 3 pipirig are within the LHD scope. He staff has not approvul the application of LDH for these piping for operating reactors, nere are differences in ASMB Code requirements betwcx Clars I and Clans 2 and 3 piping. Discuss the significance of these differeneca on ensuring piping structural integrity and describe procedures to address them.

l'or example, the ASME Code does not require a fatigue analysis for Clua 2 and 3 piping. Discusa how the fatigue resistance of the LilB Class 2 and 3 piping will be addremd. As another example, the in servic4 inspection rniuirements for Class 2 piping is based on a sampling basis and Class 3 piping is based ,n visual inspections.

Divuss any augnwnted in acryice inspection for Class 2 and 3 LilB piping,

Response

ne leak before-break snethodology is not applied to ASME Clans 3 piping as presented in response to RAI 210.6, nis methmlology is applied to Class 2 portions of the main steam and ferdwater piping inside containment, ne nusin difference between the ASME Section 111 stress analysis requirements for Class I and Class 2 piping is that there is no cumulatk %tigue damage calculation for Class 2 systems.ne fatigue resistance of the Clan 2 piping which nurts the lea. ro-break criteria is verified by perfornung fatigue crack growth calculations for postulated part through-wall th he welds selected for examination u part of the ASMB Section XI in service inspections willinclude a termins 1 weld at the s'eam cenerator nouje for one main steam line and one main feedwater line, ne terminal end locations typically hav a, er thermal and SSB streaca than other Imations and are generally limiting for leak 4cfore-break evaluatio - ne size of the sample and the frequency of inspection will not be increased beyond that required by the ASMl3 Ne.

SSAR Subsection 3.6.3.2 will be revir4xl as tollows:

SSAR Revision:

Foe-A&ME-Cod +rS*4 bin 411rCh= 3 d:dgaed 1 4 ping-end-1 4 pingw44.ignal4o-ASME Ce, Ethe Ilirthe-pipiagr *ppo'4* rend +tru 4*e4+* 4 igac44oe41 " Autdownw ohqek: w =' h wo- t servbe-enJ4*-*#evk+4napn4i"Matuiren:=' fe: ^ SMR-Cmler &*4 ion 413rClaw-2-1

. 4 pinpre-met,

• For ASMU Class 2 piping a fatigue crack growth analysis is performed to verify the fatigue resistance of ,

the piping sygem, in addition, the welds selected for enmination for the ASME Section XI in-servict inspection will include two terminal end welds to the steam generator nonles: orr for the main steam piping and one for the main feedwater piping, 252 M W Westinghouse 4

l

NRC REQUEST FOR ADD 1110NAL. lNFORMAT10N Ouestion 252.0 Sxtirm 3.6.3 of the SSAR indicates that LDti may be applied for portions of pipmg outside containment. Provide information to denamstrate the reliability, elfestiveneu, sero.itivity, and timeliness of leakage detection nwthmis and prmedures selected for outside containment.

Response

The leak before-break methodology is not applied to piping outside of c<mtainment 1his nethodology is applini to piping my items that extend into the break exclusion nme of the main steam tunnel outside c4mtsinment. This applica to the main steam and main feedwater piping as follows. The main steam piping f:om the steam generator outlet nonic to the anchor downstream of the imlation valve is analym! for applicable loadings including the SSE.

This anchor is at the exterior wall of the auxiliary building. The portion of this piping from the containment penetration Huod head intard wcld to the above anchor satisfies the break excit.sion zone requirements of Standard Review Plan 3.6.2. The portiori of this piping from the steam generator outlet nouje to flumi head inimard weld is evaluated to the leak before-brer.k methodology The main favjwater piping from the steam generator inlet nonje to the anchor upstream of the isolation valve is analyttd for applicable loadings including the SSH. 'Ihis anchor is alm hicated at the exterior wall of the auxiliary building "the portion of this piping from the containment penetration Hued head inimard weld to the above arnhor satisfies the break exclusion wne requirements or Standard Review Plan 3.6.2. The pc. tion of the piping from the steam generator inlet noule to the flued head inboard weld is evaluated to the leak before-break methodology.

SS AR Sutsections 3.6.3 and 3.6.3.1 will be revised as follows:

SSAR Revision:

(Subwtion 3.6.3, paragraph 71 HigWawgy-A&MRCode r h4swi43IrC4w*4r a r*r.441 s* 4 he-im4t *msnaldian*4**4*r gw4* **1mh4 for.+.mrdiam *ith4wk heforv4wwk+inwia A+pira e, M wwdrawilm+mtainmenWev 1 4th+waluaQm

• • +m44nued-4.sh.4*t44+m4**da.ede+mtednmen4-end4wiwde*+ny4aneh-wenee4km*4 4wa*Hhe-pawar*4km andam4*we-A+4ho g4:4*g+y*4*'**1wwtion*+f+y*lenw4w+hh4 4 tie m4 twas 4ialw+4m4" J ' -4*fy the-me64u.niati+-34 p+4*wk+ilwier 4h*4*tuitementa*>dwitwin-dim.umeJ4n4ulw.makm*-344-and44,M.+4he analysie-endbothe-of-twaulakJpip.4vptut* opt4y, liigh-energy ASMB Cale Setion III piping that is evaluated to the leak-before-break methmlology is identifialin Appendit 3D. This applica to the main steam and main fealwater pip.ag na follows. The main rfcam piping from the steam generator outlet noule to the anchor downstream of the isolation valve in analyml for applicable h>adings including the SSE. Thir anchor is at the exterior wall of the auxiliary bulkhng. The portion of this piping from the containment penetration flued head inboard weld to the above anchor satisfica the break exclusion is requirements descritu! in Subsection 3.6.2. The portion of this' piping from the steam generator outlot nonje .

flued head inboard weld is evaluated to the leak before-break methmlology. The main feedwater piping from the steam generator inlet nonje to the anchor upstrearn of the it.olation valve is analyml for applicable loadings

[ WOStingh00$0

NRC REQUEST FOR ADDITIONALINFORMATION a III W  !

including the SSF.. His anchor is ah.o located at the etterior wall of the auxiliary building. The portion of this piping from the containtnent pencaration Dued head inboard weld to the above anchor satisfies the break exclusion mne reslulterrents descrital in substion 3.6.2. Tk portion of the piping from the steam generator inlet nouje to the Oued he*J inboard weld is evaluates! to the leak-before break mellaidology. Iligh energy piping that does not natisfy the leak-before break criteria is designed to the requiremenwnts dircuacd in Subsections 3.6.1 and 3.6.2.

(Subrection 3.6.3.1, paragrsph 9) hb.lw 4ee-and44=mpipiat a 4We..untainnw44.*4h Tn4*nela*4*teiilemmininnwd4*elwedmignal 4+>4h*4*i4*i+41.vehyml4+>poviJ#4cek-bim,4 .ek4wa.4.wh44*4n high enwyptin** -

(Subwtion 3.6.3.1, paragraph 15)

Ovtaide a m6minnwil r+im+al 4dacews hen ev4*e*l4n*4 runwedat hm i+*.<44a tri. vide 4etwa km wpsvoltm64o RegulatotyGuid*4A&r -Oul*id*<Mm4minnwd-vou rJ4* ring 4.f-the-viwein-44carnam44cedwelerdinee6-6*4+ugylNewnicJ hy-pwh*J+ingw4keend4 wmidity-owe.ovenwd*4*4h++ece -e.lja.ed4+>4he4inamod4h*4* deli ++wiven,-

s 252.6 2 V,f Westinghouse

-g

NRC REQUEST FOR ADDITIONAL.lNFORMATION Ouestion 252.10 Section 3D of the SS AR discusses the LilB evaluation for the reactor coolant loop piping. The SSAR indicates that two different soil conditions have been considered in deriving piping stresses. Discuss how these piping strer.ses represent the worst condition of all potential sites within the scope of AP600 applications.

Response

As describalin Appendix 311 of the SSAR. two soil conditions for the reactor cadant loop pipe stress analyses were shoren to provide preliminary strews for a sample of the application of the leak before-break methodology.

Analyses for other soil conditions will be performed by December.1993. These loop analyses will represent the worst condition of all potes al sites within the scope of AP600 applications.

SSAR Revision: NONE 3 W85tlngh0US8

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 252.11 Tables 3B-3 and 311-4 of the $$AR give strews used in the LBB evaluation of the reactor coolant imp piping.

Provide information to clarify whether the strenes are frorn the strer.s analysis of souted or unrouted reactor coolant hop piping.

Response

The stress analysis for the reactor coolant piping is included in the $$AR Appendix 3B to provide a 6arnple for the application of tne leak-before-break rnethodology. This analysis is bami on routed reactor coolant piping which is supported by the prirnary equipnwnt supports; the connecting piping (e.g. surgeline)is not included in the nulel.

SSAR Revision: NONE W Westinghouse l

w. _. -, - - _ - . _ .

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 252.47 Discuss whether the carbon content of austenitic stainle6s steel in the reactor intemals and core sup;mrt sinactures will be limited to less than 0.02 % as recommendal in NURE04313. Revision 2 'Tahnical Regert on Material Selection and Processir2g Guidelines for BWR Coolant Pressure Boundary Piping,' January 1988. Provide a technical discussion on why this limit is not relevant to the proposed use if it is not used (Section 4.5.2).

Response

ne carbon content of austenitic stainters steel will not be limited to less than 0.02 % C in the reactor internals and core support structures. NUREG-0313 in concerned with noiling Water Reactors where presence of oxygen, from radiolytic decomposition of water in the core, can lead to 10 SCC. EPRI NP-6780-L recommends a limit of 0.035 %

as suitable for Preuurized Water Reactors where the presence of oxygen is suppiest.e4 by the hydrogen overpressure. The EPRI industry guidance will be followed where L or LN grades are to be used.

SSAR Revision: NONE 2s2m W westingnouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 252.60 Section 5.2.3.4.3 of the SSAR inc', cates that there nay b iinaccessible cavities or chambers in the RCP11. Discuss considerations to eliminate these ;onditions. If these c4 3ditions cannot be avoided, provide accest.cs for future inwrvice inspection to monitor ia conditions in these c4 vities or chambers. Diwuss the awociated augmented inservice inspection program.

Response

The discussion of inacceuible cavities and chs.nbers is relnud to the requirement for a test for sensitiation for stainless steel parts. Those components that tre not simple shapes are subject to the test. A configuration that precludes rapid cooling of a part when water quenched necessarily preclude or complicate the subsequent inservice inspection of the pressure boundary. As noted in 5.2.4.2, the Clau 1 components can be inspected per the requitements of the ASME Code,Section XI. The first paragraph of 5.2.3.4.3 will be revised to reflect this response as follows:

SSAR Revision:

Austenitic stainless steel naterials of product forms with simple shapea need not be corrosion tested provid-ed that the solution heat treatment is followed by water quenching. Simple shapes are defined as plates, sheets, bars, pipe, and tubes, as well as forgings, fittings, and other shaped products that do not have inaccessible cavities or chambers that would preclude rapid cooling when water-quenched. This characteriution of cavities or chamters as inaccessible is in relation to the entry of water during quenching and is not a determination of the component accessibility for inservice inspection.

[ WC5tiflgt10USB I

NRC hEQUEST FOR AD0lil0NALINFORMATION Ouestion 252.107 1he proposed new steam generator tube plugging criteria in Section 5.4.2 of the SSAR would place increased emphasis for stearn generator integrity on prirnary to secondary leakage nxmitoring relying on increnal sensitivity and on line real time read <>uta. Describe Westinghouse's propowd plans on implementing this nxmitoring.

Response

1he tube repair criteria and inservice inspection program is the resp (msibility of the Combined Ikense holder. The SS AR does not delineate a tube repair criteria that would rely on leak monitoring. The inervice inspection progtsm is expected to follow the industry guidelines for sicam generator tube inspection contained in the EPRI steam generator examination guidelines report. Please atm see RAI Questions 250.10 and 250.21 for a discunion of the inservice inspection program. See response to RAI 250.19 for a discuuion of the development of tube repair criteria.

The AP600 has radiation monitoring capability in the steam generator blowdown, rnmin steam lines, and conderu.cr air removal discharge to detect radiation due to primary to secondary side leakage. The radiation nxmitoring capability for the secondary system in the AP600 is in conformance with industry requirenwnts for advanced light water reactors.1he nxmitoring system is in conformance with the recommendations in NRC Information Notices 88-99 and 91-43 to provide for the detection of increases in primary-to-secondary leak rate. The radiation monitoring system is diwussed in Subsection i1.5 of the SSAR.

SSAR Response: NONE 2s2.10 w 1 E Westinghouse

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I NRC REQUEST FOR ADDITIONALINFORMATION Question 252.108 Descrite how the ' Delta-75' steam generator design proposed for the AP600 will facilitate the implenwntation of in-situ fusion techniques for steam generator tube repair. Alwi, discuss how the selection of noterials for the tube support structures and the tubesheet will prtwiude deleterious effects on rnaterial toughness cauwd by in-situ fusion heat efftwts (Section 5.4.2).

Response

The steam generator design features provide for tube repair with autornated equipnent without manned entry. The repair equipment and the robotic delivery equipment that has ten developal for use in operating steam generators can be used in the AP600 steam generator. Please also see the response to RAI 250.15. The steam generator uses materials for the tubesheet and tube supporta plates that have teen used in previous steam generators in the same applications. With respect to the use of repair techniques the material r. election and design of the steam generator for the AP600 is the same as or similar to previous s. team generator designs. Specific testing of the effects on tubesheet and tube support plate nuiterials to support the application of in-situ fusion techniques for steam generator tule repair in the AP600 steam generator is not required.

See RAI 250.15 for suggested SSAR changes related to provisions for the me of robotic equipruent for steam generator repair.

SSAR Revision: NONE 252A084 i W~

Westinghouse l

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 252.109 Section 5.4.2.3.3 of the SSAR indicates that tule vibration has p>tential to cause wear. Discuss in detail the p>tential for wear degradation with emphasis on the AP600 features that are designed to mitigate this concern.

Response

The potential for tube wear is dependent on the amplitude of vibration of the tube, the material couple, and the configuration and surface condition of the supports being impacted by the tubes. The arnplitude of vibration is determined, in part, by the sire of the gaps between the supports and the spacing between supports. The AP600 steam generator includes a number of features that minimize the potential for tube wear at tube supports and antivibration bars. Provisions to minimize the potential for wear include the spacing between the tube supports, the configuration of the broached hole through the support plate, the clearance lxtween the tube and the hole in the tube support plate, tube support plate material selection, and the configuration of the anti vibration bar assemblica.

The seventh paragraph of SSAR Subsection S.4.2,3.3 will be revised to reflect this responw as follows:

SSAR Revision:

Tube vibration response is shown to have wear potential within available design margins even for limiting tube fit-up conditions, based on pervious experience in fabricating steam generators with lit-up control typical of the AP600 steam generator. The AP600 steam generator includes a number of features that minimize the potential for tute wear at tube supports and antivibration bars. Provisions to minimize the potential for wear include the spacing between the tube supports, the configuration of the broached hole through the support plate, the surface finidi of the broached hole in thMute support plate, the clearance between the tube and the hole in the tule support plate, tube support plate material selection, and the c4mfiguration of the anti-vibration bar assemblica.

Cone pondiog-Tube bending stresses corresponding to tube vibration rispmse remain more than two orders of magnitude below fatigue limits as a consequence of vibration amplitudes constrained by available clearances.

These analyses and tests for limiting pntulated fit up conditions include simultaneous contributions from flow turbulence.

W Westinghouse

NRC REQUEST FOR AODITIONALINFOllMATION Ouestion 252.111 Recent plant operating experience disclosed the ponibility of miss-placed anti-vibration bars (AVDs) and the possible revere conuquences. Discuss how the proper location of AVBs will be ensured (Section 5.4.2).

Response

The potential for misplaced anti-vibration bars is minimited by an in-process dimensional inspection of the tube U-bends and the anti-vibration bars. As discussed in NRC Bulletin 88-02, mir.placed anti-vibration bars have resulted in adverse consequences only in conjunction with dented or corrosion packed top tube support plates. De AP600 steam generator is not expected to be subject to these conditions due to the design and material selection of the support plates.

The last paragraph of SSAR Sul+ection 5.4.2.3.3 will be revised to reflect this response as follows:

SSAR Revision:

The U-bend fatigue (discussed in NRC Dulletin 88-02)is not a consideration in the AP600 steam genera-tors. The mechanism considered in Dulletin 8842 requires denting of the top tube support plate. But this is not expected with the stainless steel tube support plates in the AP600 steam generator. Additionally, the location of anti-vibration bars is controlkd by in process dirnensional inspection.

2s2.,,,,

W westingnouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 252.112 Industry recommendations and other vendors' improved steam generators designs incorporate prinary side manways having a minimum inner diameter of 53 cm (21 in). Discuu Westinghou e's technical basis for limiting the posts in the ' Delta-75' steam generator to 46 cm (18 in) in diameter as indicated in Section 5.4.2.5 of the SSAR,

Response

The function of primary manways is to provide access to the channel head for inspection and repair, if necessary.

The steam generator provides full access to the tubes with automated, robotically delivered inspection and repair equipment. Use of this equipment minimizes or eliminates manned entry into the channel heeJ for routine inspection and repair activies. The 18 inch diameter manway is large enough for use of the inspection and repair equipment.

A larger diameter manway opening would increase the radiation levels in the areas adjacent to the manway when open. 'Ihe 16 inch manways in operating steam generators have been found to be adequate, if not optimum, to permit entry into the charmel head. The 18 inch diameter manway size is based on considerations that include accean, space available, stress levels in the channel head, size of closure fasteners, occupational radiation exposure, and handling of covers.

SSAR Revision: NONE 2s2.u 2a w wesunsouse I

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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 252.113 lixperience has shown the advisability of complete records and archive materials to investigate corrosion and mechanical damage which may occur during service. Industry recommendations suggest archiving at least 2 m (7 ft)of each heat of row I and 2 'U bends' prior to final heat treatment and following the mill anneal, and production samples containing tubes from each heat expanded in a tube sheet rrockup. Archive samples should be maintained to support future chemical cleaning programn and for pouible defxt calibration samples for inservice inspection.

Describe WestinghouWs program to retain records and archive materials (Section 5.4.2).

Response

The heat and not of tubing material for each steam generator tube is recorded and documented. Archive samplea of each heat and tot of steam generator tubing material are provided to the combined license holder for use in future materials testing programs or inservice inspection calibration standards. The archive samples are subject to the same manufacturing processes and inspections as the installed tubing. A mmimum of seven feet of tubing in the final heat treat condition is supplied.

The archive samples to be provided do not include tubes expandal into tubesheet mockups. The archive samples provided are consistent with the ALWR utility requirements for advanced light water reactors and EPRI guidelines for PWR steam generator tubing specifications (EpRI Report NP 6743.L) which do not include a requirement for tubesheet mockup archive samples. Tubesheet mockup samples have generally not been provided with replacement steam generators.

The following paragraph will be added to SSAR Subsection 5.4.2.4.1 to reflect this response:

SSAR Revision:

The heat and tot of tubing material for each steam generator tube is recorded and documented as part of the quality anurance records. Archive samples of each boat and lot of steam generator tubing material are provided to the combined license holder for use in future materials testing programs or as inservice inspection calibration standards. A ruinimum of seven feet of tubing in the final heat treat condition is supplied.

W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 252.114 Provide detailed discussion on the extensive operating esperience and laboratory testing (including nundel boiler tests) to justify the une of all volatile treatenent (AVT) r.econdary u ater chemistry with inconel 690 for the proposed 60-year plant design life (Section 5.4.2).

Response

Battground information on the material and corrosion properties of nickel <hromium-iron Alloy 690 may be found in EPRI report NP-6997 M, ' Alloy 690 for Steam Generator Tubing Applications.' The report is a compilation of published and previously unpublished data on the testing of Alloy 690 and includes information on corrosion behavior and a comparative ranking of Alloy 690 with nickel-chromium-iron Alloy 600 and nickel iron <hromium Alloy 800. In this ranking, Alloy 690 was found to provide either comparable or additional corrosion resistance for a wide variety of postulated steam generator crevice environments relative to the other candidate tubing materials. The report concluded that Alloy 690 is 'the material of choice for steam generator tubing applications tweause of its corrosion resistance in a variety of environments.*

The industry has also recently completed a review of secondary water operating chemistry guidelines from the viewpoint of limiting secondary-side initiated conosion concerns. The review, presented in " Interim PWR Secondary Water Chemistry Recommendations for IGA / SCC Control," EPRI TR 101230, September, 1992, reaffinns the use AVT secondary water chemistry in plants which 1 ave not experienced corrosion. This recommendation does not preclude the possibility of adopting an alternate water chemistry as technology evolves, nor does conformance to the guidelines assure that steam generator tubing integrity will be maintained for the 60-year operating design objective. Nevertheless, selection of Alloy 690 tubing and adherence to the AVT water chemistry guidelines provide reasonable assurance for maintaining the long-term integrity of the steam generator tuben under current technology assumptions, in addition to the water chemistry and the tube alloy material selection, steam generator design features have an affect on the tube integrity of operating rteam generators. Currently, there are 106 operating steam generators utilizing, to some degree, AP-600 type design features. After apprmimately 700 cumulative years of operating experience with Alloy 690 and Alloy 600 tubes, not one tube has been removed from service due to secondary-side initistal corrosion.

SSAR Revision. NONE W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 252.115 Address the potential for primary water stress corrosion cracking in Inconel 690 for die proposed 60-year plant design life (Section 5.4.2).

Response

The resistance to prinwry water stress corrosion cracking (PWSCC) of thermally treated nkkel<hromium-iron Alloy 690, chosen for the steam generator heat transfer tubing, has been the subject of extensive corrosion testing in simulatrd and highly accelerated primary side environments. These corrosion test programs were conductrd individually and in joint programa by reactor vendors and primary metals suppliers in the U.S., France Sweden and Japan over note than a ten year perimi.

The major results of thew evaluations have been summarimd in a rwent EPRI report (NP4997). The unaninous cont.ensus of these efforts is that Alloy 690 appears to be highly resistant to PWSCC at the temperatures and operating conditions appropriate to the plication in steam generators. See the respcmse for RAI 252.114 for additional background of the tube material selection. Periodic inservice inspection of the steam generator tubing affords an opgxntunity to see that the integrity of the steam generator tubing is mainta;ned for continued operation of the steam generators.

SSAR Revision: NONE 252.115 1 1

3 Westinghouse l

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t NRC REQUEST FOR ADDITIONAL. lNFORMATION

{

Ouestion 252.110 Address the resistance to corrosion of inconel 690 in upset w ster chemistry conditions which would take place over the proposed 60 year plant design life (Section 5,4,2).

Response

As discussed in the response to RAI 232.114, nickel <hromium-iron Alloy 690 tubing provides comparable or additional corrosion resistance over a wide range of postulated crevice environrnents relative to other tested

candidate tubing materials, nese environments bracket those which may be formed as a result of operation under upset water chemistry conditions.

Tube material wiection and design fr.atures alone will not assure nor guarantee that the steam tenerator tube lategrity will be maintained for the 60 year operating design objective of the steam generator without defects or significant degradation. Periodic inservice inspection of the steam generator tubing affords an opportunity to see that the integrity of the steam generator tubing is maintained for continued operation of the steam generators.

SSAR Revision: NONE 3 Westinghouse

NRC HCOVEST FOR ADDITIONALINFORMATION Ouestion 262.137 Identify the Steam and fmlwater sye. tem naterials and provide infornution to denonstrate that the materials meet the requirements of Sntion lit of the A$ Mil Code ($ntion 10.3.6).

Response

Subsection 10.3.6 sprifies that nuterial nelection for ASMB Cmle, Sxtion 111 Clau 2 and 1 components in the main steam and feedwater systena are addrened in Subuwtion 6.1.1.1. where commitment is made to complisnee with the ASME cale.

'The nutcrial trecifications for preuure-retaining materials in each component of an engincend safety featurca system meet the rtyuirements of Article NC 2000 of the ASME Code, Section 111 Class 2, for Quality Group D; Article ND-2000 of the ASMll Code, Section 111, Clain 3, for Quality Group C .ompmentst and Article NE 2000 of the ASMli Cale, Section til for containment pressure imundary components.*

The ASME Swtion til portion of the main steam line is currently spe:ilied as cathon steel A/SA-106 Grade D, whereas the indwater line is currently opwilial an alloy steel A/SA 355 Orade F22. The SSAR commitment in that the materials und for these lines will comply with the ASME Code as cited above. Material 6 election could be affxtal by ongoing piping analys.es and leak before break qualification analyses for these linea.

SSAR Revision: NONE 2 s2.i a7.,

w wes,ineouse

NRC REQUEST FOR ADDITIONAL.lNFORMATION Ouestion 262.143 Although the condensate p>lishing system serves no safety related function, show that failure of any of its compments will not cauw damage to the systems in{uired for safe plant shutdown (Section 10.4.6.1.1). j

Response

As doncribed in Subsection 10.4.6 of the Ap600 SS AR, the nonsafety related c<mdensate polishing system is a partial flow system that operates in parallel with inmin condensate system flow. Condensato polishing flow can be isolated if necessary during p>wer operation. 'The condenute polishing system piping and components are located in the turbine building and are completely isolated from the t.afety related equipment required for safe shutdown of the plant. The systems raiuired for safe Shutdown are listed in Section 7.4 of the AP600 SSAR and their successful operation is not irnpacted by the f ailure of comp >nents in the condensate polishing system.

SSAR Re, vision: NONE 252.143 1 3 Westinghouse l

l NRC REQUEST FOR ADDlil0NAl INFORMATION l

i Question 420.5 nere is no di4cussion on annunciator system and the guidance for manual action in $ntion 7.5 of the SSAR. Dere are no analyses to address defense-in depth design to protect against the comnxm mode failures in the Integrated protection System and the Integrated Control Systern. Provide such analysea.

Response

1) Annunciator system and Ouldance for manual action.

De AP600 annunciator system is described in Section 18.9.2 of the SSAR. De following clarincation describes the way expected operator resixmse to plant alarms will be addren.ed in the context of the Ap600 Man Machine Interface.

The AP600 M MIS utilites the wall panel infonnation station for display of high level alarm indications. Once an alarm is activated, the operator can retrieve information related to the alarm from the operational displays at the operator's conele and can query the alarm system. He appropriate alarm response procedure can be called up at the operator workstation and appropriate controls can be acceased. This operating philosophy is expected to be applied to all operational nules in which the alarm system remains functional. In the event of a loss of the alarm system, the Class IE qualined display pnxessing system is used.

The alarms are alm embaldal in the operational displays. Thus if the wall panel information station is not available, the infonnation available on the wall panel information station is alm available at the operator's console if an operator's console is inactive, the supervisor's console has the informatica available, and the plant control functions can be switchal over to the superviwr's console.

If the operators' consoles. the supervisor's console, and the wall panel information station are not available, the operator moven 'o the center bridge console between the operators' consoles to use the qualified display processing system and A alicated hard controls, guided by the emergency operating pmcedures, to bring the plant to safe shutdown anu .o maintain that condition. The operator can me paper alarm response pmcedures and the dedicated controls, with fuxlhack from the qualined display processing system, to maintain the plant in a safe condition. If any alarms are identified through the function based task analysis that prompt a requirni operator action to bring the plant to a safe shutdown condition and maintain it, they will be pmvided in both the qualifini display processing system and in the operational display systern.

Particular guidance conceming when to take manual action is provided through the alarm response proculuren, as well as the operating procedures (normal, abnormal, emergency). These are derived from the output of the function basal task analysis, after the task allocation between human and computer has been made.

W Westinghouse

NRC REOUEST FOR ADDITIONAL INFORMAll0N

2) Analysis of defenw in depth de$ign.

A defenne indepth analysis of the protation and $sfety nxmitoring system, as described in NUREG-0493, is currently beir,g perforned, and will be subrmital upon completion.

In the caw of a comnum nxxle failure in the plant control system, the protection and safety nxmitoring system will still be available, and will provide protection for the plant. In the case of a comtnon mode failure that afIccts both the protection and a,sfety oxmitoring and plant control systema, the diverse actuation system will be available to protect the plant. Comtmm nuede failures of the protation and safety trxmitoring system and diverne actuation system were includal in the probabili6 tic Ri6k Asunament (PRA), therefore, a defenne-in-depth and diversity muentrent is not required for the plant control system.

SSAR Revision: NONE i

420.5 2 Vj Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 420.6 The AP600 design implemented an integrated control system (ICS) as indicated on Figure 7.1 1 of Section 7.7 of the SSAR. Ilowever, there is no discussion in Section 7.7 to describe the ICS, no analysis to address protection against common nule failutea in the ICS. Provide this infmmation.

Response

The integrated control system is implemented by the Plant Control System described in Section 7.1.3 of the AP600

$$AR.

The analysis to protect oganst common nale failure in the Plant Control System was done as part of the Probabilistic Risk Assesstnent (PR A). The specific fault trees developed to address common cause failures of the plant control system are:

CLCCX - Comnon cause failures of control logic cabinet.

CMUXCCX Common cause failures of control multiplexer cabinet COCCX + Common cause failures of control group cabinets SEGSELCX Conunon came failures of signal selector subsystems in the PRA, failures of the Plant Control System, including common cause failures, were analyred together with failures, including common cause failures, of the Protection and Safety Monitoring System and the Diverse Actuation System. This analysis of the AP600 instrumentation and control systems is dercribed in the PRA in Appervlix C20, ' Protection and Safety Monitoring System, Plant Control System', Appendix Cl2, ' Diverse Actuation System', and Appendix U.3.4.6, ' Evaluation of Common Cause Failure for Instrumentation and C<mtrol.'

The conclusion is that the nonsafety-related Plant Control System, taken together with the r.afety-related Protection and Safety Monitoring System and the nonsafety-related Diverse Actuation System is, by PRA analysis, sufficient to protect the plant for the analyral events.

SSAR Revision: NONE W Westinghouse

NRC REOVEST FOR ADDITIONAL. lNFORMATION Ouestion 435.5 1he non-safety ac p>wer systems nuy fait during a seismic event or fire, in which caw, the installed non r.afety ac power systena nuy not be available beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Soc tion 8.3.1.1.1 of the SSAR states that a provision of two 480 V non< lass 1H transportable diesel generators (150 KW cach) is nude to uni the po.t 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> power mjuirements following extended loa of all electric po tcr iources. Discuss the provisions for ensuring the availability of the transputable diesel generator to the safety system.

Response

Two transportable, ac diesel generators willbe .unnected directly by prefabricated cables to the Class IU regulating transforners, hydrogen recombmers, and temporary equiprnent tiansported to the site as listed on SSAR Figure 8.3.1-4. 'ihese two tutnsportable diesel generators will be stored at a location far enough from the site no that they renuin unaffected by events such as earthquake and explosions and will be transportable to the site within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

1he storage and transpntation details are covered by the combined license applicant.

i SSAR Table 1.8-1 will changed to include the following:

SSAR Revision:

8.9 Storage of transportable NNS Combined License 8.3 AC diesel generators applicant coordination WO51lRgIl0USC

NRC REQUEST FOR ADDITIONALINFORMATION K

Ouestion 435.7 Section 8.3.1.1.2.1 of the SSAR states that each standby diesel generator is dedicated to one of the two divisions of permanent non safety loads. Describe whether the two non-safety ac divisions will tneet the physical and clutrical independence requirernents of RO 1.75.

Response

SS AR Table 8.1 1 provides the information on the Regulatory Guides and their applicability to the electrical systems design. As per Table 8.1 1, RO 1.75 is not applicable to the SSAR Subsection 8.3.1 which addresses the non-safety ac power supply system design. The nonsafety related ac power is for invesment protection only and is not required to meet the electrical indepemience requirement of RG 1.75 The folowing statements provide the electrical separation aspects of the nonufety-related ac power supply system design:

• Non Class IB circuits are electrically isolatn! from the Class IE circuits in compliance with the RO 1.75 stipulations, and IEEE Std. 384 rniuitenwnts.

• Nonsafety-related ac power system design includes two divisions of permanent nonsafety-related loads each supported by its own onsite diesel generator unit.

• Each onsite diesel generator unit is hicated in a separate enclosure.

• The ac switchgear units pertaining to each of the nonsafety related ac divisions are h>cated in separate roona in the Annen lluilding.

• The control power for the control of the nonsafety-relatal ac switchgear breakers is provided from t.eparate nonsafety relatal de power sources.

• The detailed raceway and circuit design for the two non-safety ac divisions is yet to be completed.

Regulatory Guide 1.75 separation criteria is not going to be implemented for the nonsafety-related me raceway design.

SSAR Revision: NONE l

l l 435.7 1 l

T Westinghouse l

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NRC REOUEST FOR ADDl110NAL INFORMATION Ouestion 435.9 Periodic testing and test loading of a diesel generator in a nuclear power plant is a nxcuary function to demonstrate the operability, capability, and availability of the unit on denund. Periodic testing couptnt with good preventive naintenance practices will awure optimum equipment endiness and availability on loss of offsite power. To achieve this optimurn equipment traliness, the following itemn should be considered:

a. The equipment should be tested with a minimum huding of 25 percent of rated load. No load or light load operation will cause incomplete combustion of fuel and the conwquences could le the potential equipment failure due to the rum a..d varnish depaits and fire in the engine exhaust system.

b. Periodic surveillance testing should be performal in accordance with the recommendations of the engine nanufacturer.

Provide a discussion of the naintenance and testing program for the diesel generators.

Rr+pmse:

A diwussion of the propaxi maintenance and testing program for the diesel generators is providal below:

Maintenance The diesel generators are not safety related and will be maintained in accordance with the ruluirements of the overall plant maintenance program. This program will cover the preventive, corrwtive and predictive maintenance activities of the plant systems and equipment and will be presented in the combined license application.

Perimlic Testing

. The periodic testing program will be performal in accordance with the recommendation of the engine manufacturers.

The specific engine loading level will be determined bued on the engine manufacturer's recommendations such that the load operation will not cause incomplete fuel combustion that may result in gum and varnish deposits in the engine exhaust system.

SSAR Revision: NONE

[ Westingh0tlSe

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 435.12 i l

Provide infonnation on the maximurn loading of each supply circuit during nornal and ebnornal operating  !

conditions, including accident conditiona and plant shutdown conditions, to demomtrate the adequacy of the lines, circuit breakers, and transformern of the offsite power system.

Responso:

The design of the offnite power system is a site ymcific issue. The information requested can not be determined until the physical characteristics of the offsite power system Lave been designed. This infornation will be determined by the Combined License applicant during site selection and design of the tran5miaxion system. For furt er infornation, pleaw refer to the reyonse for RAI 435.4.

SSAR RevlSlon: NONE l

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l 435.12 1 3 Westinghouse

NHC REQUEST FOR ADDlil0NAl. lNFORMATION Ouestion 435.17

• the frnquency of catastrophic failures of the main step up tramformers due to lightning has boca greater than what was anticipated Describe the design for lightnir.g protection of the main step-up transformers.

Response

The main step-up transformers art protected from lijtning coming from two nources. IJghtning can affect tramformers by both a direct strike to the transformer and also by lightning propagating to the transformer over the transminion lines connecht to it. Therefore, two means of protection are ut.ed. Grounded shield wires are located above the equipment in the transformer area, including the main step-up transformers, to intercept lightning strikes in the area and conduct them to ground. Suitably rated surge arresters are located on the high voltage side of the main step-up tramformers to mluce the magnitudes of incoming lightning caused voltage surges to levels which are well within the insulation withstand capability of the transformer.

SSAH Revision: NONE 45"'

w wesunsouse

1 NHC REQUEST l'OR ADDITIONAL INFORMATION Ouestion 435.24 Provide the details of the design of the de power system that anures equiprneat will be protected from damaging overvoltages from the battery chargers that may occur due to faulty regulation or operator error, j l

Response

The battery chargers are provided with a liigh DC Voltage Charger Shutdown feature, which includes an alarm relay and indicating light The liigh DC Voltage Charger Shutdown functionally disables and locks-out the battery charger whenever the output de voltage exeeeds the preset upper limit of charging voltage. DC equipment and components are rated for snaximum equalitation voltage of 140V, Dui, the de system cquipment is protected from overvoltage damages.

SSAR Revision: NONE

[ W0stinghouse

NRC REOUEST FOR ADDITIONAL INFORMATION J

l.i.! "II!

n Question 435.25 Section 8.3.2.2 of the SSAR states that the Clan IE de system is ungrounded; thun, a single ground fault does not cause immediate loss of the faulted system. Ilowever, a ground fault followed by a snond ground can produce ground currents of sufficient nugnitude to initiate operation of de energiud de loads or inhibit drop 4>ut of energiud de loads. Detation with alarms is provided for each division of power so that ground faults can be located and removed before a ucond fault could disable the affected circuit, This has been the subject of Information Notice 88-86 and 88 86, Supplement 1. Describe the ground detection system for the ungrounded de auxiliary system.

Response

During the detail design phase, the concerns expressed in Information Notice 88 86 and 88-86, Supplement I will be considered for specifying the Class IE de ground detection system. Also, plant proculures will be established so that prompt action is taken to clear any ground fault on the Class IE de system, SSAR Revision: NONE W Westinghouse

NHC REQUEST FOR ADDITIONALINFORMATION L

Question 435.32 Describe how the Class IE power systems meet RG 1.75.

C

Response

Compliance with RG 1.75 is discussed in Section 1 A of the SSAR. The AP600 Class 1E power system design will meet the physical independence requirement of RG.1.75 as described in detail in Sections 8.3.2.3 and 8.3.2.4 of the SSAR. AP600 design will be baml on IEFE Standard 3841981 'IEEE Standard Criteria for independence of Class IE Equipment and Circuits *, A brief outline of the AP600 design is presented below:

The separation of safety-related system will be maintained by the physical layout and the unique identification of equipment, raceways and cables. Different color codes will be used for different safety-related electrical equipment, power, control and instrumentation cables and raceways, in addition, each circuit and cable will carry a unique identification number which will provide a means of distinguishing between circuits of different separation divisions.

Cables associated with the same division will require no physical separation c . may be routed in common raceway.

Signal cables associated with the same division will require no physical separation and may be routed in common raceways.

Class IE cables and electrical and 1&C equipment (panels, cabinets, components, etc.) will be identified with the appropriate division identifier (A, B, C, or D),

c Raceways and cables associated with a specific division (A, B, C, or D) will be physically separated from all other divisions, and from non-safety raceways and cables, as described in SSAR Section 8.3.2.4.

The indeperdence of non-Class IE circuita from Class IE circuits will be achieved by complying with the requirements addressed in IEEE 384-1981, paragraph 5.6.

Physical separation between electrical equipment and components acc.aciated with redundant divisions will be consistent with the criteria established in IEEE 384-1981, Sections 5 & 6. The Class IE equipment of each division will be located in safety-related structures.

SSAR Revision: NONE

[ Westingt100S8

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 435.33 Show that the design of the de power system will ensure the reliable operation of the system's loads and equipment for the full range of operating voltages, including charging, equalizing, and end-of-discharge. The analyses should include the steady state and switching transients.

Response

'The AP600 de power system design is based on 'IEEE Recommended Practice for the Design of Safety-Related DC Auxiliary Power Systems for Nuclear Power Generating Stations -IEEE Std 946-1985". The batteries have been sized in accordance with IEEE Std 485-1983. The battery duty cycles (load profiles) used for sizing the batteries are shown in Tables 8.3.2 - 1,2,3 & 4 of the SSAR. These duty cycles include steady state and switching transients as indicated by 0-1 min,1-1440 min, and 14404320 min kW power requirements. The number of plates selected for each cell is based on battery manufacturer's curve for end-of-discharge cycle voltage of 1.75V/ cell, i.e.,

battery endef-discharge voltage of 105V, Also, aging factor (1,25), temperature factor (1.064) for minimum operating temperature of 67'F, and design margin of 1.1, as recommended by the IEEE Standard 485, have been umi to calculate the number of plates. The selected batteries and battery charger sizes and the Class IE de power distribution configurations are shown on SSAR Figure 8.3.2-1 (Sheets I and 2).

The operating voltage range under all operating modes, including charging, equalizing, and end-of-discharge is 105 to 140Vdc. The maximum equalizing charge voltage for batteries is 140Vdc and the end-of-discharEe volt *Se i8 105Vdc. The nominal system voltage is 125Vdc.

'Ihe operating voltage range for the equipment and the associated loads will be specified in accordance with Table 1 of IEEE Std 9461985 to ensure reliable operation of the de power system for the full range of operating voltages, including charging, equalizing and end-of-discharge. Also, as described above and in the response to RAI Question 435.28, the batteries have been sized to provide steady state and switching transient power within the required voltage range of 105 to 140Vdc.

SSAR Revision: NONE W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION y T.I.

Question 435.65 Section 8.3.2.2 of the SSAR states that the Class IE battery chargers and Class IE transformers have built-in circuit breakers at the input and output sides for protection and isolation. In some cases, they are serving as isolation devices between Class IE and uon-Class IE circuits. This is not in accordance with Position C.1 of RG 1.75, which precludes the use of interrupting devices actuated only by fault current as acceptable devices for isolating non-Class IE circuits from Class IE or associated circuits. Provide acceptable isolation for the Class IE circuits in accordance with Position C.1 of RG 1.75, or provide justification for deviation from this position.

Response

Conformance with RG 1.75 is addressed in SSAR Section 1 A. The purpose of the built-in circuit breakers provided at the input and output sides of the Class IE battery chargers and the Class IB regulating transformers is for the protection of the Class 1B equipment against a fault, in addition, the input side breakers equipped with shunt trip devices will be used to isolate the Class IE battery chargers and the Class IE regulating transformers in the event of out of-tolerance power outputs from the non-Class IE sources. These breakers are actuated not only by fault current and therefore, may be considered as isolation devices. However, the pnmary isolation functions are performed by the Class IE battery chargers and the Class IE regulating transformers.

The Class 1E battery chargers receive 480V non-Class IE ac input power through isolation transformera which are built into the battery charger. In addition, the battery chargers are provided with blocking diodes to prevent accidental discharge of the batteries in the event of a fault at the non-Class IE side of the battery chargers.

l The backup ac power to Class IE UPS loads is provided through 480-208/120V Class IE regulating transformers.

In this mode of operation, the inverter / battery will be disconnected from the UPS system. The regulating transformers are provided with isolation transformers and power conditioners which protect the Class IE 120Vac system from browuouts, high line voltages, surges, and electrical noise.

In view of the above, the CLss IE battery chargers and the Class IE regulatmg transformers together with the built-in circuit breakers meet the isolation device requirements in accordance with RG 1,75, in addition, IEEE Std 384-1981, Section 7.1.2.3 lists inverters, regulating transformers, and battery chargers as an acceptable isolation device.

SSAR Revision: NONE

[ W85tiflgfl0USB

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 435.68 Section 8.3.2.1.1.2 of the SSAR states that if an inverter in the UPS system is inoperable or the Class IE 125 Vdc input to the inverter is unavailable, the power to the 120 Vac bus is transferred automatically to the backup ac wurce, ne backup ac power supply la Figure 8.3.1.1 (Sheet 2 of 2) of the SSAR is down to be a non-Class IE rource. Traditionally, this ac source has been classified as a Class IE source in existing plants and in the evolutionary designs. Where is the Class 1E/non-Class 1E boundary? Does it allow connection of the transportable diesel generator through the Class IE system only? The regulating transfonner assembly feeding the Class IE battery charger should be classified as Class IE. Revise the SSAR sccordingly, or providejustification for not doing so.

Response

The Class IE and non-Class IE boundary is at the regulating transformer. This makes the backup ac source to the inverter Class 1E, ne transportable diesel generators are connected through the Class 1E system only. As indicated by Note 2 on Figure 8.3.2-2 of the SSAR, the regulating transformers are Class IE qualified. Also, it is further clarified in Section 8.3.2.1.1.2 of the SSAR which i,tates 'The backup power is received from the diesel generator backed non-Class IE 480Vac bus through the Class IE regulating transformer. A reference may be made to the responses to RAI 435.5 and RAI 435.65 for further clarifications.

SSAR Revision: NONE W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION

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..i Question 435.69 Identify the vital areas and hazardous areas where emergency lighting is needed for safe shutdown of the reactor and the evacuation of perr.onnel in the event of an accident. Describe the lighting provided to accommodate those areas so identified.

Response

As described in Section 9.5.3.2.2 of the SSAR, the emergency lighting will be provided in the vital areas and hazardous areas as follows:

The main control room and remote shutdown areas have been identified as vital areas where unergency lighting is required. Lighting in these areas consists of 120 Vac fluorescent lighting fixtures which are supplied from the Class IE dc & UPS system. There are no other areas required to achieve safe shutdown of the reactor.

Emergency lighting in plant areas for safe ingress and egress of personnel following loss of all ac power is provided with sealed beam fixtures. He fixtures have self contained battery with eight hour rating and battery charger powered from both onsite and offsite AC power. Power to these units automatically switch to their internal de source once normal ac power is lost. Ingress / egress routes are used for evacuation of personnel in the event of an accident. See also question RAI 435.72.

SSAR Revision: NONE u s.6 M W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 440.22 Various computer codes have been referenced in Chapter 4 of the SSAR, most of which have leen used to analyze dif? ent Linds of Westinghouse fuels, i.e., the STD 17X17, OFA and VANTAGE 5. Although a cursory review of chapter 4 indicates that there are great similarities between the AP600 fuel and previously designed Westinghouse fuel, minor dilferences do exist.

n. Can these previously approved codes accommodate / address the differences that exist between the AP600 fuel denign and earlier Westinghouse designs?

b. Does sufficient data (experimental or otherwise) exist to demonstrate the capabilities of these established codes to handle any AP600 differences 7

c. Describe the data base that justifies the DNB correlation and limits applied to the AP600 fuel, i

Response

Parts a. and b.

With regard to the codes used in Chapter 4 for nuclear design, thermal-hydraulic analyses and fuel design, the previously approved cales referenced in this Chapter can accommodate the very minor differences in the AP600 fuel design. Relative to the previously approved VANTAGE Sil design (Section 4.1, Reference 1), the only  ;

changes in the AP600 fuel design within the active fuel region of the fuel assembly is the addition of one low pressure drop structural grid and one IFM grid. The neutronic, thermal-hydraulic and mechanical design characteristics of the AP600 fuel design are essentially identical to that of previous proven designs. Sufficient experimental and operating data exists which has been used in the qualificatica of the nuclear design, thermal-hydraulic design, mechanical design, and fuel desiF u codes to handle minor differences in the AP600 design.

Part c.

1he WRil 2 correlation will be used to analyze the AP600 fuel. The database for this correlation is described in WCAP-10444-P-A. The AP600 fuel geometry is within the licensed parameter range of the WRB-2 correlation.

The flow rates at the time of minimum DNBR of the loss of flow event are somewhat below the previously licensed WRB.2 lower limit on flow (G = 0.9 x 108 lb/ft -hr). 2 DNB testing is being done to verify the extension of the WRB 2 correlation to these lower flows. See the response for RAI 440.1 for information en the DNB testing. The W 3 correlatie will be used where the WRB-2 correlation is not apphcable, e.g., steamline break.

SSAR Revison: None

[ W85tingl100S8

NRC REQUEST FOR ADDITIONAL INFORMATION N

Ouestion 440.23 l

Section 5.4.7.2.1 of Chapter 5 of the SSAR provides "a summary of the specific AP600 design features that address j SECY 88-017 regarding mid loop operations." l

a. The staff believes that the reference to "SECY 88417' stated in this section is intended to mean ' Generic Letter (GL) 88-17.* Please confirm.

b. In addition to GL 88-17, the staff also addressed its concerns regarding shutdown and low power operations in NUREG 1410 and provided guidance in draft NUREG-1449. Are the AP600 design and the proposed Technical Specifications consistent with the findings and recommended industry actions discussed in NUREG 1410 and draft NUREG-14497

c. The staff is aware that many issues related to shutdown and low-power operations are the plant owner's responsibility as they are related to operation, maintenance and refueling plan, procedures and risk management. What are the specific guidance or requirements that must be implemented by the AP600 plant owners to fulfill their responsibilities with regard to shutdown or low power operations?

d. Section 5.4.7.7 of the SSAR states that the r tor-operated valves connected to the hot leg are interixked to prevent them from opening when RCS p. essure exceeds 450 psig, and to prevent their being opened unless the isolation valve from the IRWST to the NRHRS pump suction header is closed. Confirm that there is no autoclosure interlock for the AP600 NRHRS isolation valves that could result in loss of decay beat removal due to unplanned activation.

e. Are ther. any systems or components needed for shutdown cooling which are deenergized or have power locked out during plant operation? If so, disciiss what actions have to be taken to restore operability to the components or systems, and describe where tne actions must be taken.

Response

a. The reference to *SECY 88-017* stated in this section is intended to mean

• Generic Letter (GL) 88-17."

The SSAR will be updated to reflect this change.

b. In addition to the design features described m Section 5.4.7.2.1 of Chapter 5 of the SSAR which provided "a summary of the specific AP600 design features that address GL 88-017 regarding mid-loop operations,"

the AP600 contains passive safety grade protection against a loss of normal RHR during shutdown. The recommendations of the specified NUREGs are incorporated into the design and technical specifications of the passive r;afety systems. During mid-loop operations, the IRWST is operable, and the ADS valves connected to the pressurizer are open. If the normal RHR system is lost, the IRWST will provide core cooling by gravity injection. Technical Specification 3.5.4 addresses this issue.

Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION

c. He combined license applicant is responsible to ensure that the AP600 design requirements, system operating proceduren, and maintenance and testing recommendations are properly implemented during plant operations, including specific requirements related to shutdown, refueling, and low power operation. The requirements for shutdown operations are containcxlin the AP600 Technical Specifications. Please see the responne to RAI 440.28 for a description of the requirements during shutdown operations.

d. There is no autoclosure interlock for the AP600 NRilRS isolation valves that could result in loss of decay heat removal due to unplanned activation.

c. The normal RilR system contains valves that isolate the system from the reactor coolant system hot leg (RNS-V001 A,B RNS-V002 A,B) have power locked out at the motor control center during plant operation.

Prior to initiating normal RilR cooling, the operator will be required to go to the motor control center to restore power to the valves. Note that this will occur prior to mid-loop operation.

SSAR Revision:

5.4.7.2.1 Design Features Addressing Mid-loop Operations The following is a summary of the specific AP600 design features that address SECY RS 09 Generic Letter (GL) 88-17 regarding mid-loop operations.

440.23 2 W Westingttouse

NRC REQUEST FOR ADDITIONAL INFORMATION 1

Question 440.30 The AP600 design does not fully comply with the staff position stated in SECY-90415 with regard to interfacing system LOCA. For example, it is stated in Section 1.9.5.1 of the SSAR that the normal residual heat removal system pump seal is not designed for the higher NRilRS pipir.g design pressure and could fail if overpressurized.

Though the AP600 design of NRilRS limits the leakage to within the capacity of the chemical and volume control system to allow the plant to be place in safe shutdown, the leakage from the pump seal will bypass the containment.

As stated in the draft Commission paper entitled " Issues Pertaining to Evolutionary and Passive Ught Water Reactors and Their Relationship to Current Regulatory Requirements," (Letter from D.M. Crutchfield to E.E.

Kintner dated February 27,1992), the staff concludes that, to the extent practicable, all elements of the low pressure system, including pump neals, should be designed to withstand the full RCS pressure. For those systems and components that do not meet this requirement, justifications must be provided as to why it is not practicable to do so,

s. Provide a list of systems and compments interfacing with the RCS that are not designed to have ultimate rupture strength at least equal to the full RCS pressure,

b. For each of these systems or compments in item (a), either (1) confirm that it will be redesigned to meet the ultimate rupture strength criteria, or (2) providejustification as to why it is not practicable to meet this criterion, and confirm that these systems or components willinclude the capability for isolation valve leak testing, valve position indication, and include h:gh pressure alarms as discussed in the draft Commission paper.

Response

a. The AP600 design features that address intersystem LOCA are described in SSAR section 5.4.7.2.2. He AP600 position with regards to this issue is contained in SSAR section 1.9.5. The normal residual heat removal system is designed to the extent practible, to have an ultimate rupture strength equal to the full RCS operating pressure. The normal RilR pump seal is the only component or system connected to the RCS that is not designed to have an ultimate rupture strength equal to the normal RCS operating pressure.

b. The following is a discussion of the practicability of designing the RIlR pump seal to full RCS operating pressure.

The rupture pressure of the RilR pump mechanical seal was investigated with suppliers of nuclear grade mecht.nical seals. The rupture pressure of the mechanical seals in the AP600 RilR pumps is approximately 670 to 700 psig with the pump operating. The rupture pressure is somewhat higher when the pump is not openating.

There is a fundamental problem with designing an RIIR pump seal that can withstand full RCS pressure. Any type of seal that can withstand the full RCS pressure will likely have abnormally fast wear of the seal faces during normal plant operation at low seal pressures. This increased wear at normal plant operating conditions could well prevent the seal from maintaining the pressure boundary if ever exposed to the full RCS pressure. Use of the high W Westinghouse i

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k NRC REQUEST FOR ADDITIONAL. INFORMATION - _

t-J l l pressure seals will also require more frequent maintenance during normal operation. In simple terms, the seal can be designed for normal low pressure operation or for full RCS pressure, but it is unlikely that a seal can be designed to operate desirably at both conditions.

It is the AP600 position that the best solution is to utilize the existing single seal design in conjunction with a sturdy _ ,

~

disaster buaing design. This combination will maximize the reliability of the seal during normal RHR operation .

. and minimize maintenance and associated radiation exposure. Furthermore, this design approach will minimize the leakage from the normal RilR pump in the event that the mechanical seal fails. The leakage from this design can e be controlled so that only a small portion of the water that leaks past the primary seal faces escapes to the pump cubicle and the majority of the leakage is piped to a controlled drain. This is more favorable than a seal specially; designed for full RCS pressure at the expense of normal condition reliability, Because of the design features of the AP600 which prevent the likliehood of exposing the normal RHR system to -

full RCS pressure, and because of the limited leakage from the normal RHR pump seal, the current PRA analyses '

has shown that the interfacing system LOCA accident has been eliminated (contribution to CM F < 103. Therefore, the current- AP600 design approach should be retained.

3 In addition, provisions for leak detection and position indication for the isolation valves that separate the normal -

RHR system from the RCS are consistent with those reconunmded by the NRC, Furthermore, high pressure alarms..

will be added to the normal RHR pump suctica to comply with the NRC position.

The following paragraph will be added to SSAR Subsection 5.4,7,2.2:

k SSAR Revision:

RCS Pressure Indication and High Alarm - The AP600 Normal RilR system contains an instrumentation channel that indicates pressure in each RHR pump suction line. A high pressure alarm is provided in the main control room --

a to alert the operator to a condition of rising RCS pressure that could eventually exceed the design pressure .of the L. normal RHR system.

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= 440.30-2 W Westingflouse

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NRC REQUEST FOR ADDITIONAL INFORMATION I

N..!" sfe air-operated valves and this failure probability is 7.lE-4.

The failure probability of containment heat removal .y. tuns in a typical Westinghouse plant is: 1) Failure of 2 out of 2 fan cooler units is 4.E-3 (assuming that all wpport systems are available) and 2) Failure of 2 out of 2 containment water recirculation systems is about 2.E-3 for the equipment. Note that operator actica is required, in a typical plant, to align the recirculation system and to align component cooling water to the residual heat removal system heat exchangers. Either one containment fan cooler or one RilR heat exchanger operating in recirculation is typically sufficient to preclude containment failure.

A very important advantage of the AP600 PCS, not reflected by a comparison of containment heat removal systems failure probabilities, is that PCS failure in the AP600 is independent of the failures that cause core damage. The only commonality is the actuation logic but in the AP600 design, this failure requires failure of the plant monitoring system. the diverse actuation system, and hard-wired manual actuation, which uses information provided by diverse indication from the diverse actuation system.

Failure of the water cooling system for the out de uf the containment shell, which results in only air cooling beir.g available to cool the containment, is the worst PCS failure evaluated in the PRA. Even in this case, sufficient cooling from only air maintains the containment pressure below Service Level C. It could be postulated that air cooling of the containment could be reduced by blocking the air flow path through the annulus with water accumulating at the bottom of the annulus if the annulus drains were plugged. He probability of such a scenario is negligibly small and the failure of the annulus drains is independent of any failures that lead to core damage.

Provisions to preclude blockage of these drains are discussed in response to RAI 720.26.

The value of containment ultimate pressure and the bases for this value are provided in SSAR Subsection 3.8.2.4,

" Design and Analysis Procedures" (for the containment vessel).

SSAR Revision: NONE 48 "

W westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION i

Question 720.3 The AP600 design incorporates a number of novel features (e.g. new controls) and passive devices. The PRA is not capable of portraying a realistic risk profile for the AP600 design unless a comprehensive and systematic search was conducted to assure that adverse failure modes of these new features were integrated into the PRA models.

By adverse failure modes, we do not only mean failures that result in challenging safety systems or obstruct the shutdown process, but also those failures that may motivate the operators to take wrong actions. Provide information detailing your efforts in searching for these failure modes.

Response

A comprehensive and systematic search was conducted for adverse failure modes of both novel features of the AP600 and of other features which are also found in current plants. The systematic strategy that was used to identify potential errors of commission is described in Section D.1,

• Errors of Commission', of the PRA report.

The first part of this approach, which consista of identification of the systems and of the modification of their operating status with potentiM influence of the successful performance of a safety function, was used to also identify hardware spurious failures that could aggravate the accident progression.

PRA Revision: NONE 220.3.,

w wesuncouse l

1 NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 720.6 he AP600 PRA does not provide a concise sumnury of sequences in which recovery actions were credited. These recovery actions are important to the staff's understanding of system performance and required operator response.

Provide a listing of top accident sequences in which recovery was credited and characterize these actions. By characterizing thew actions, the staff means identifying error probability, locations, describing the performance shaping factors that were used, listing the control room information motivating the recovery action, etc. Was credit for recovery handled before or after truncation, and was recovery integrated in the analyses at the cutset or sequence level 7

Response

For events evaluated at power, the dominant sequences are shown in Appendix F, Table F-7, ne operator recovery actions for these events are identified as LPM-M AN01, LPM-M ANO3, LPM-M AN04 ATW-MANO3, and REG-M AN00. and shown in the following 41 sequences: 5, 6, 12, 16, 18, 23, 24, 25, 28, 30, 32, 34, 35, 37, 38, 42, 53, 56, 5 8, 60, 62, 63, 65, 66, 68, 69, 70, 74, 75, 76, 78, 80, 81, 82, 83, 89, 91, 93, 96, 97 and 100.

For events evaluated at shutdown, the dominant sequences are shown in Appendix F, Table F-10. He operator recovery actions for these events are identified as IWN-M AN00, CCN-M AN02, RilN-M AN03, and PRN-M AN02S, and shown in the following 21 sequences: 11, 16, 17, 18, 19, 30, 31, 34, 35, 48, 40, 02, 66, 70, 80, 81, 84, 86, 88,94 and 95, For events evaluated in the containmant analysis, the dominant sequences are shown in Appendix G. Tabie G 5.

The operator recovery actions for these events are identified as CIX-M AN00, SFN-M AN00, and ZON-M AN02, and shown in the following 33 sequences: 1, 4, 11, 16, 26, 33, 36, 38, 45, 64, 76, 100, 112, 113, 132, 136, 138, 139,151,163,168,169,172,175,176,179,180,181,185,189,190,258 and 259.

Table 720.6-1 provides the recovery actions, the descriptions, the expected cues for the operators, the performance shaping factors used in the evaluation, the location where the action is performed (that is; in control room or hically), and the estimated human error probability. Two recovery actions, REG-MAN 00 and ZON-MAN 02 are performed locally; the other recovery actions are carried out in the control room. The TiiERP methodology was used to evaluated the recovery actions, except CIX-M AN00 and ZON-MAN 02; the HEPs for these two actions are based solely on engineeringjudgement, and are believed to be conservative.

The recovery actions were integrated into the PRA at the fault tree or cutset level. Therefore, credit for recovery was handled before truncation, PRA revisions: NONE W Westingt100se

NRC REOUEST FOR ADDITIONAL INFORMATION Table 720.6-1 Recovery Actions idenufier Operator Action Cues l PSF Location IIEP Note LPM-M AN01 Failure to recognize the low SO wide r*:.ge long procedure; Control 2.20E43 I need for reaetor coolant 6:d, ingn hos leg Tinw window (Tw) room system depressuritation temperature; low hot = 30 min; high during a small loss of leg w ster level stress level conlant accident or loss of high-presauro heat removal system LPM M ANO3 Failun to recognize the low hot leg wster long procedure; Control 8.30E42 2 need for resetor coolant level, Tw = 15 min; room system depressuriation jammed instrument high stress level uhen only the diverse setuation sy stem is providing information during a small loss of coolant acciJent or transient LPM .M ANO4 Failure to recognue the Low hot leg wster long procedure; control R.30E42 2 need for resetor woolant level; Tw = 15 min; room aystem depressurization jammed instrument high stresa level when only the diverse actuation system is providing infonnation during a medium loss of coolant accident ATW M ANO3 Failure to recognize the law narrow range Short procedure; Control 1.53 E42 3 need and failure to level in steam Tw = 1 min; nmm manually trip the reaetor generators; high moderate stress within I minute, given pressunzer pressure; level anlicipated transient flow mismatch without seram between feedwater flow and turbine inlet pressure ItEG-MAN 00 Failure to regulate the Estimated 7 to 10 Short procedure; local 2.10E-Ol 4 stanup feedwater different alarms Tw = 50 min; following full opening of indieating loss of high stress level the regulating valves compressed air after a loss of compressed air 1

1 720.6-2 W_ WestinEhouse

NRC REOUEST FOR ADDITIONAL INFORMATION Table 720.6 Recovery Actions -_

IJennifier Operstor Action Cues PSF La etion liEP Note IWN MAN 9) Failure to manually open I gnmp of alarma long procedure; Control 11$E43 $

two rm> tor-operated indicat ing failure of Tw = 60 min; room valves during shutdown the nortnal residual high stress icvel conditions with the heat remtwel system; normal residual heat low hot leg wnier renovn! system level unavailable CCN MAN 02 failure la exclude heat liigh temperature on long procedure; Control 2.62002 (i exchanger the line downstream Tw = 60 min; nom 11001 A and align 11001B of the heat exchanger moderate stress during normal operation level RilN MANO3 Failure to recognize the Voltage return on the Short procedure; Control 1.97D03 7 need and gnd, Tw = 120 min; room failure to manually no now from the high stress level restart the normal normal residual heat residual heat renwval removal system; low mystem pumps following hot leg temperature grid recovery within two hours aRer a loss of offsite power and failure of both automatic and manual transfer onto a die nel generator, during shutdow n PRN-M AN025 failure to actuate the RilR pumps trip; RCS 1 ong procedure; Control 2.55 E03 8 )

passive residual heat pressure increase (up Tw = 30 min; room removal system (PRilR) ta normal RilR relief high stress level air operated valves, valves opening V108 A and V108B setpoint); low following loss of offsite pressurizer level; low power dunng shutdown hot leg w ater level SFN-M AN00 Failure to recognite the Low IRWST level; long procedure; Control 9.40E04 9 need and failure to low reactor cavity Tw > 120 min; room perform IRWST makeup level high stress level with spent fuel or themical and volume control system aRer containtnent isolation following gravity injection 720.6-3 Wa West,nEh0USBi

NRC REQUEST FOR ADDITIONALINFORMATION Table 720 6-1 Recovery Actions IJentifier Operator Astion Cues PSF lastion llEP Note CIX MAN 00 raiture to deteet loss of less of plant power N/A; engirwering Control 1.00E41 10 aU instrument stion ard generation judgement applied nom control, and diverse indication system slarms ZON M AN02 Failure to start diesel loss of AC power N/A; erigineering lual 1.00E-02 11 gencrutor lmally in order with kiss of renxas judgenwns applied to prmiJe power for operation capability long term RCS makeup following loss of control nom control lM-720.6 4 W -

Westingh0US0 4

NRC REQUEST FOR ADDITIONAL INFORMATION NOTES

1. The LPM-M AN01 operator action represents diagnosis of the event to perform actions that could be associated with three systems or subsystems, (the ADS, RCS and CMT). Here are cues that are unique to actions for each system, as well as cues that are common to particular cases for the different systems. The PRA models considered the cases in which common cues are provided to the operators for the different system failures.

These cases are believed to be representative of all possible cases.

In general, the human reliability analysis considers the presence of the reactor operator and the senior reactor operator throughout an emergency. A moderate dependency is assumed between the operators. Recovery is applied for the presence of the shift supervisor and shift technical advisor (STA) for time windows greater than 10 minutes.

2. The LPM-MANO3 operator action represents diagnosis of the event when only the diverse actuation system is providing information. It is believed that the signals for this event will be generated much later than the signals for the event LPM-MAN 01, discussed above. It is estimated that a time window of 15 minutes exists for recognizing the diagnostic cue for this event. No credit is taken for the presence of STA in the analysis, due to the lack of instrumentation and the relatively short time window.

The factors that are considered in the evaluation of LPM-M AN03 are applied to LPM-MAN 04.

3. The ADWMANO3 operator action to trip the reactor has a time window of I minute. No recovery credit is applied in the evaluation of this event, due to the I minute time window.

4. The REG-M AN00 operator action is perfornwd locally. The flow must be regulated by stroking a manual valve, which is considered to be relatively difficult to perform. No credit is taken for the presence of the STA in the evaluation of this kical action.

5. The IWN-M AN00 operator action is carried out within a short procedure, and is required to be completed in an estinuted time window of I hour. Slack time is believed to exist; therefore, recovery credit is applied for the presence of the shift supervisor and STA.

6. The CCN-MAN 02 operator action is carried out within a somewhat lengthy process. The estimated time window for this task is about I hour. Slack time is believed to exist; therefore, recovery credit is applied for the presence of the shift supervisor and STA. -

7. The RIIN-M ANO3 operator action is carried out within a short procedure, and is required to be completed in an estimated time window of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Slack time is believed to exist; therefore, recovery credit is applied for the presence of the shift supemsor and STA.

8. He PRN-M AN02S operator action is carried out within a long procedure, llowever, there are few steps to be completed in an estimated time window of 30 minutes. Slt.ek time is believed to exist; therefore, recovery credit is applied for the presence of the shift supervisor and STA.

W Westinghouse

NRC FtEdUEST FOR ADDITIONAL INFORMATION l

9. The SFN-M ANDO operator action is carried out within a long procedure, and is required to be completed in an estimated time window greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Slack time is believed to exist; therefore, recovery credit is applied for the presence of the shift supervisor and STA.

10. 'Ite CIX-M AN00 operator action considers the loss of plant power generation. For this scenario, plant instrumentation is not providing information to the operators. it is estimated that the most limiting time window for detecting this failure is 40 minutes for a " cold leg break' accident; a time window of about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is estimated for the transient event. The shorter time window is reflected in the IIEP of 1.00E-Ol; this value is believed to be conservative.

11. The ZON-MAN 02 operator action is a local action to start the diesel generator. The time window to complete this task is estimated to be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No other performance shaping factors are considered in the evaluation.

The IIEP of 1.00E-02 is assigned as a screening value based on engineering judgement.

720.6-6 W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION f

Question 720.7 Provide information describing how cutsets were truncated and how the truncation limits were selected.

Response

The cutset generation was donc by the WESCUT/WESLGE codes on a IBM-PC 486 with a software limit of 1000/2000 cutsets per fault tree. Following this the fault tree linking process was done. This process consists of first a

• REDUCTION" (linking subtree cutsets into a system cutset file) of system cutset files, then the accident sequence quantification(multiplicationof fault trees making up accident sequences). The WLINK code system was used for fault tree linking. This code system has a limit of 9900 cutsets on the PC and 100,000 cutsets on the UNIX workstation. Most of the " REDUCTION" wa done on the IBM PC, some of the reduction and all of accident sequence quantification where done on a UNIX workstation, where software limits on the number of cutsets are larger.

A. For fault tree quantification, a 1.0 E-09 cutoff probability was used. When this cutoff caused the software to go out of limits, one of the following two actions were taken:

1. For highly redundant systems, such as ADS, the original fault tree was broken into artificial subtrees, which were individually quantified and later their cutsets were linked.

2. For other fault trees, the cutoff probability was increased until the software ran and provided an ample number of cutsets representing the system failure.

B. For fault tree " REDUCTION" the objective was to use a cutoff probability of 1.0 E-12, to the extent possible. if the software limit for number of cutsets was exceeded, the cutoff probability was increased until the software nm, and an ample number of cutsets representing the system failure were obtained.

For some larger system cutset files, the reduction was done on a UNIX workstation where a low cutoff probability can be chosen, since the software limit for number of cutsets is 100,000 on the Unix wmkstation.

C. For event tree sequence quantification by fault tree linking, cutoff probabilities ranging from E-10 to E-15 were used, based on the initiatiag event frequency. For lower initiating event frequencies, a lower cutoff value was used to allow system level linking to occur at least at the E-10 level or lower.

Based on the care provided in this process, and the involvement of various cognizant engineers in review of the output cutset files, there is confidence that all dominant cutsets were captured in the output files.

D. Precautions were also taken in construction of the fault trees to avoid unnecessarily challenging the number of cutset limits of the software. This is done as follows:

T Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

! I In construction of the fault trees, implied OR logic in constructing some basic events was used, thus reducing the number of independent basic events to a minimum. (* implied OR logic", combines as many independent failure nxxtes of a component as possible, into a single basic event.) When the number of ba. sic events are reduced to a minimum set in the fault tree, the resulting number of cutsets representing the fault tree is also minimized, and the failure probability of each cutset is increased (in effect, sone cutsets now represent "fanulies of cuttets'),

This modeling process does two things:

1. It allows representing the same system with a lower number of cutsets (avoids exceeding software cutset limits);

2. It allown picking up more cutsets at a given cutoff probability (because more cutsets now have higher probabilities since they actually represent families of cutsets).

An excessive number of cutsets was not required to adequately represent a fault tree. The dominant core damage cutsets were captured using this process.

PRA Revision: NONE 720.7-2 W

- WestinEhouse

NRC REQUEST FOR ADDITIONAL. INFORMATION Ouestion 720.10 Several of the core damage sequences result from the failure of the NRIIR or CVS to inject coolant into the vessel.

These systems must be aligned for injection, and these actions result in containment bypass if the pumps fail or suction is not available. The stalf is concerned with the potential to bypass the containment from failure of the NRilR or CVS when it is aligned to inject coolant into the vessel during LOCA or transient sequences that lead to core damage. In order to prevent containment bypass, the operators must isolate the system from the reactor vessel when these systems fail to inject. These operator actions were not evaluated by Westinghouse. Describe the hunum actions that are planned to isolate NRilR and/or CVS upon failure to inject and how these associated human error probabilities will be quantified.

Response

Injection from the RNS and the CVS are taken credit for in the AP600 PRA. The potential to bypass the containment due to failure of the RNS or the CVS does not increase because of the following:

CVS - The system can only inject water from tanks located outside of the containment; it does not have suction connections from the IRWST or the RCS. Note that water can be removed from the RCS by the CVS, but water cannot be recirculated through the CVS, as in current plants. Instead it would be sent to the waste processing system. Following actuation of an *S* signal, the RCS letdown and the connection to the waste processing system are isolated and these connections are not reopened to provide RCS injection from the CVS.

The process of aligning the CVS to provide injection does not increase the chance of a containment bypass because all of the valves between the CVS makeup pumps and the RCS are normally open except one check valve. In addition there are eight isolation valves in this path between the RCS and the makeup pumps.

These isolation valves include simple check valves, fail open air operated valves, de powered motor operated valves and stop check valves. With the redundancy and diversity of valve types, in particular the different types of check valves (simple and stop) the probability of a containment bypass occurring on failure of injection would not contribute to the frequ;ncy of release from the plant.

RNS- The RNS can inject water into the RCS from the IRWST located inside the containment. Failure of the RNS to inject is almost entirely due to either failure of the operators to start the system, failures of the pumps to start, or failures of MOVs to open. None of these failures lead to containment bypass because either the RNS would remain isolated from the containment or, if the RNS was aligned and the pumps failed to start, the pressure boundary of the RNS would still be intact. Note that since the RNS pressure boundary is Regulatory Guide 1.26, quality group C / seismic category 1 and it is designed for 900 psig, the chance of its rupturing and causing containment bypass is insignificant. Also, note that the RNS does not have connections that are used to transfer water to or from other systems outside of containment.

720.10-1 l W~~

WestingflouSe l

t 1

NRC REQUEST FOR ADDITIONAL INFORMATION R

As discuned above, it is unnecenary for the operators to take actions to prevent containment bypass, when they attempt to use the CVS or the RNS to provide RCS injection and these systems fail. As a result, the AlYaOO PRA does not consider operator action to isolate these systems from the RCS or the containment.

PRA Revision: NONE W"

20.10-2 W Westingflouse

l l

l NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 720.16

'Ihe staf f observes that the t.evere accident initiators for the external events analyses for the Al%00 design were either screened out based on qualitative analyses or they had unusually low frequencies when compared with similar sequences for the current generation of operating plants. Provide a listing and characterization of AP600 design features that have a major contribution to reducing risk from external events to the level estimated in the AP600 PRA, as well a comparison between the major assumptions in the analyses and those generally used in published PRAs.

Response

a. Al%00 design features which reduce risk:

al. The Al%00 is a passive plant. Major contributors to plant risk in many previous PRAs have been the failures of pumps to provide decay heat removal. Hecause the Al%00 does not rely on active components, such as pumps, to remove decay heat, fire induced failure mechanisms of active components are not present. Active components are credited in the Al%00 PRA, but decay heat removal can be achieved without their success, a2. The Al%00 incorporates spatial separation in its layout, which prevents a single event from disabling a function. For instance, of the six vital battery rooms, four are located on one elevation in separate rooms, while the other two are located on a different elevation in separate rooms. There is also a spare battery room in addition to the six mentioned above, which is located in a separate room.

a3. Most of the safety systems which are required to reach a safe stable state are located inside of the containment, and are not subject to external event effects.

b. Comparison between major assumptions used in AP600 PRA and existing PRAs bl. Internal Gooding:

i) Assumptions which were used in the Ai%00 internal flooding PRA were similar to those used for flooding PRAs of present day plants. In addition, flooding hazards during reduced power and shutdown operations were considered for the .AlbOO PRA. Most existing PRAs consider at-power hazards only, consistent with the requirements of NUREG-1335.

b2. Internal fire:

i) Harrier failure was not considered credible in the AP600 fire PRA, whereas it is considered in NUREG/CR-4840. Harrier failure is not considered in other methodologies approved by the NRC (e.g.,

NUREG/CR-2300, Fire Induced Vulnerability Evaluation). The combustible loading in areas separated 720.1 G-1 WM Westinghouse I

L

NRC REQUEST FOR ADDITIONAL INFORMATION iu

. ie, by barriers is low, so barrier integrity will not be challenged by a maximum fire. Moreover, all barriers proposed for the AP600 are designed based on NFPA standards. For these reasons, barrier failure was not considered to be credible.

(ii) Offsite power was assumed to be available during a fire event. Appendix R analyses for existing plants assumed a loss-of-offsite power concurrent with a nre. The AP600 PRA, however, considered whether a fire could cause a loss-of+ffsite-power, and it was concluded that this was not a likely event.

(iii) Containment fires were not quantitatively analyzed for the AP600 PRA. Oil spills from reactor coolant pumps have historically been the source of containmeat fires. The " canned motor pump" design of the AP600 reactor coolant pumps prevents oil spills and nres, in addition, the AP60C containment does not have any confined spaces where a damaging hot gas layer could form. Due to the absence of risk initiators, containment fires were not considered in the PRA.

Please note that the other external events like high winds, tornadoes beyond the AP600 design basis as well as transportation and nearby facility accidents, etc. are screened out per the guidelines described in NUREG-1407 on the basis of hazard frequency being acceptably low (less than 1 x 10 6 per year). Such an assurance of low hazard frequency is obtained during the selection of the site and complying with the site selection criteria as described in the SSAR.

PRA Revision: NONE 720.16-2 W - Westinohouse a

r NHO REQUEST FOR ADDITIONAt. lNFORMATION W n7

}

Question 720.23 lt is stated on page 2-4 of the PRA that air cooling of the containnwnt is suf ficient to maintain containment presure below the yield stress, yet the succesa uiteria for the passive cooling system (page C7 2) requires operation of one of two branches of the water supply system. Remlve thin apparent inconsistency.

Response

he sucte*n criteria discusm! in section C 7 describes the success criteria for the design basis of the PCS system.

Ilowever, it is true that air cooling of the containment is sufficient to prevent the cetainment from reaching the yield stree.s and from faihng. %e operation of the panive containnent cooling system water has a significant effect on the containment pressure and is required for design buis accidents to keep the containment preuure below the design precure. When it is act operating, the long term containment presure exceeds the design preuure, but remains below the containment ultimate capacity. De PCS node is on the containment event tree becauw the containment precure, and therefore the PCS water operation, impacts two important downstream event tree results:

1. The success enteria for the ex-vessel debris coolability node is dif ferent when the PCS water is operational than the success criteria when the PCS water is not operations!. Debris is considered to be coolable in the AP600 cavity when the PCS water is on and 3 out of 4 water sources (core makeup tanks and accumulators) or in-containment refueling water storage tank (IRWST) water have been injected into the reactor cavity, if PCS water is not operationa),4 out of 4 water sources or 1RWST water are required to maintain water coverage of the ex-venel debris. The difference is due to the amount cf water that is prewnt as steam in the preasurital containment atnesphere, and is therefore unavailable to be in the cavity pool cooling debris.

2. The thsion product release from a containment for which the PCS water has maintained the long-term prenure below the design preuure is auigned to the OK release category. The release from a containment z in which the long-term prenure exceeds the design pressure due to a failure of the PCS water is assigned to the OKP release category. The difference is due to the increased leakaye from a presuriral containment. Since the offsite boundary doses for the release criteria are no small (I rem and 25 rem),

the difference between design basis leakage and presurized leakage needs to be addreswd in the containment event tree.

DRA Revision: NONE 5

f 72o.23.,

W wesunsouse

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 720.20 Provide additional information regarding the drains in the annular reFi on, including: (1) the normal position, and means of actuation of any valves in the lines, (2) the sire, nurnber, and ratal flow capacity of the lines, and (3) provisions for preventing bhx.Lage of the sump and linea.

Response

Two annulus drains are locatal at low points in the upper annulus fkor; the first kwated in the north uxtor of the shield building and the mond kicated in the south mtor. The drain sump in covered by a coarse screen to preclude large mateiials from bkwking the pipe inlet. Each drain is always open (without imlation valves) to the storm drain system and cital to accept maximum PCS system flow as uming a full passive containment cooling water storage tank without consideration of any evaporation and wit's a mini.num (1msling of the annulus required. Additionally, if the storm drains are bhwked, e.g. frorrn, the interface between the annulus drain and the storm drain system is an open connection such that the annulus drain would simply overflow the connection into a catch basin or into a ground level storm drain to assure positive drainage.

SSAR Subsection 6.2.2.2.4 will be nvalifial as follows:

SSAR Revision:

The cooling water not evaporatest from the vessel wall flows dowc to the lottom of the inner containment annulus into floor drains, ne drains route the excess water to storm drains. The drain line is always open (without isolation valves) and is sital to accept natimum passive containment cooling system flow. The interface with the storm drain system is an open connection such that any bhxkage in the storm drains would result in the annulus drains overflowing the connection.

W-Westinghouse

NRC Flf 0UEST FOR ADDITIONAL INFORMATION Ouestion 72o.30 De review of the AP600 PRA indicates that M AAP 4.0 was used in the analysis, it is of concern to the staff that the code was used without support from more detailed txchanistic analysis for items that are controlling containnent performance. In this regard, provide additionalinformation regarding any supporting analyses perfornxxl to confirm the adequacy of the MAAP treatnwnt of the following items: (1) external cooling of the reactor venel, (2) temperature induced hot leg failure, (3) mode of reactor venel lower head failure (creep rupture versus local failure), (4) carly containment challenges, such as direct containment heating, and fuel coolant interactions, (5) hydrogen combustion, (6) coolability of core debris in the cavity, (7) fission pr(xluct doc <mtamination factors, and (8) molten core concrete attack and non condensible gas generati<m.

Response

(1) ne M AAP 4.0 rnodels calculate that the lower head of the reactor vessel m;i not fail when it is covered by water in the reactor cavity. Supimrting analysea for the external cooling of the venel lower head can be found in WCAP 13388, 'AP600 Phenomenological Evaluation Summarica', section 2, 'A Phenonwnological Evaluation Summary on External Cooling of the RPV in Support of the AP600 Risk Aucument'.

(2) ne MAAP 4.0 analynes assumed that a hot leg failure would occur when the hot leg metal temperature exceeded 1100'K at an RCS preuure of 2500 psi. His criterion was determined in an analysis that is described in greater detail in the response to RAI 720.35.

(3)(4) M AAP 4.0 auumes a local creep failure of the vessel that is ablated as the debris is ejected. De initial radius of the failure is a user input. Because of the depresurization system, and hot leg temperature induced failure, veuel failures occur at low pressure, and the results are not very sensitive to vessel failure nule, liigh preuure nnte of reactor venel failure and direct containment heating are addrened in WCAP 13388, section 3,

'A Phenomenological Evaluation Summary on liigh Pressure Melt Ejection and Direct Containment lleating in Support of the AP600 Risk Assenment'.

(4) M AAP 4.0 anumes critical heat flux (CHF) quenching of debris in-vennel and ex veuel. Supporting adiabatic quench calculations show that the peak preuure from quench is well below the ultimate pressere of the containment.

Explosive fuel coolant interaction is addrened in WCAP-13388, section 1, 'A Phenomenological Evaluation Summary on Steam Explosions in Support of the AP600 Risk Aucument'.

(5) M AAP 4.0 treats hydrogen combustion using the same models as M AAP 3.08. The cale contains a hydrogen-air steam flammability curve, which is used to determine the time at which a tale may become flammable.

Supporting hydrogen mixing and combustion analyses are presented in more detail in PRA report chaptera 14, and 15, eppendices N and 0, and WCAP-13388, section 5, 'A Phenomenological Evaluation Summary on the Probability -

and Consequences of Deflagration and Detonation of Ilydrogen in Support of the AP600 Risk Auenment*,

(6)(B) MAAP 4.0 treats debris coolability and molten core concrete interaction using the same models as MAAP 3,00. Supporting analyses of debris coolability, core concrete interaction ar.d non-condensible gas generation are 22o.30.,

w weSunsouse

NRC REQUEST FOR ADDITIONALINFORMATION addressed in WCAP 13388, section 4, 'A Phenonwnological Evaluation Sununary on Molten Core-Concrete Interaction in support of the AP600 Risk Assessrnent.

(7) M AAP 4.0 treats fission product transport and deposition with the 6arne nondimensional empirical correlati(m that were developed for M AAP 3.08. Supporting analyses for fission product decontamination factors are presented in WCAP 13388, 'A Phenomenological Evaluation Summary on Fission Product Retention Capability in Support of the AP600 Ri6k Anneunwnt'.

PRA Revision: NONE 720.30 2 3 Westingh00$6

1 I

l NHC REQUfST FOR ADDITIONAL INr0HMATION Ouestion 720.35 Describe and justify the nulels, input muumptions, and failure criteria uml in the AP600 hot leg creep rapture aneuments. Discun the influence of the AP600 irnprovements in vessel and piping reliability on these calculations.

Response

In order to determine the reactor coolant system creep rupture criteria for the AP600, plant specific gunnetric information of the RCS piping and the Larson. Miller paranwters (ref.1) of 316 stainless steel are used to generate

' time at temperature" creep rupture curves for the hot leg nouje, cold * , nor21e, surge line and steam generator tubes. Although the steam generator tutra are inconel 690, not stainlew steel, the rupture curves are conservative because the ineonel has significantly greater strength than steel at high temperatures (ref. 2).

The analysis shows that the hot leg was the trust likely place for the failure to occur. At a prcuure of 2500 psia, all strerigth in the hot leg is lost at a temperatures greater than 1200'K. Creep rupture failure of the hot leg was modeled in M AAP 4.0 using a hot leg metal temperature threshold of 1100'K. When this temperature was reached, the hot leg was failed in the analysis. The plot of hot leg temperature and RV lower head temperature vs. time in the AP600 PRA does not show the temperature of the hottest of the two hot leg noules. This figure will be corrected in the document.

A major uncertainty in the temperature induced failure of the hot leg was the break sire, which was estimated low in the bar,e case to prevent the IRWST water from injecting and terminating the accident. This uncertainty was addressal in the M AAP4.0 sensitivity analy6cs.

References

1. Larson, F.R., Miller, J., " A Time-Temperature Relationship for Rupture and Creep Stress", Transactions of the American Society of Mechanical Engineers, pp. 765 775, July 1952,

2. Ilarrold, D.L., et al., "The Temperature Dependence of the Tensile Properties of Thermally Treated Alloy 690 Tubing *, Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Monterey, CA August 25-29, 1991.

PRA Figure L-97 will be replaced with the following:

PRA Revision:

W Westinghouse

.~.

NRC REQUEST FOR ADDITIONALINFORMATION Hot Leg ----- SG T u t' e -- -- - kV L ow He a d 1400.0 1?00 0 v

~ 1000 0 s=

/\

's V s l

( BDD C f/{

1,7

/

~#'- . . , , _ _

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~ ~ ~.NL__ -e,,

\

/

000.0 j

./ \ g ,_. -- --'--g g

,,,,,,, - s.-~.

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400.0 0.0 5.0 10.0 15.0 20.0 2 ') 0 30.0 Ttme [ hrs}

Case LTW RCS Piping and Vessel Temperaturet F10uro L-97 720.35 2 W- Westin house

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 720.30 In addition to ADS, the PRA credits hot leg creep rupture as a means of reducing the occurrence of reactor vessel failure at high pressure. In the event of the failure of the ADS, the operator is instructed to flocrj the cavity under certain conditions. Describe the influence of the cavity flooding feature on the creep rupture of the hot leg. Of particular interest is the estimated probability and timing of creep rupture f ailure with and without cavity flomling, and how these values compare to the probability and timing of venel failure under flomial cavity conditions.

Response

ne operator action to flood the cavity han little effect on the timing of the high temperature / pressure induced failure of the hot leg. He hot leg failure occurs during the period of rapid oxidation of the zircaloy cladding in the core (Figures L 90, L-94, and L-97). During this perim1, the core isjust beginning to melt, and little, if any, of the fuel containing decay heat has rehw;ated to the lower head of the reactor vessel (Figure L-99). Therefore, the reactor vessel lower head integrity is not jeopardiral until well after the failure of the hot leg has occurred and the reactor coolant system is depressurized. nese results are consistent with what is expected to occt.r since the temperature at which the cladding begins to rapidly oxidize (~1200'K) is much lower than be melting temperature of the uranium 4ioxide/ zirconium eutectic (~ 2500'K). If the lower head of the vessel is covered with water, then the reactor vessel will not eject debris into the containment. This position is supported in WCAP-13388, 'AP600 Phenomenological Evaluation Sumnutries", section 2, *A Phenomenological Evaluation Summary on Exterul Cwling of the RPV in support of the AP600 Risk Analysis",

in summary, in the AP600 PRA, for a high pressure RCS, in post-core uncovery conditions, the probability of a temperature inducal hot leg failure prior to vessel failure is 1.0, independent of cavity ikxxiing. He protwhility of vessel failure is 1.0 if the vessel has not been ikxnled by IRWST water in the cavity, and 0.0 if the IRWST water has not been injtste<l. This treatment is consistent with the treatment of other phenomena on the containment event tree. The phenomena probabilities are 1.0/0.0, but are based on the availability of AP600 systems which have finite failure probabilities.

PRA Revision: NONE W Westinghouse

3 NRC REQUEST FOR ADDITIONAL. lNFORMATION m, , 1

!! j I Question 720.38 ne staff is concerned with the twdential for fuel-ca>lant interactions (FCle) in the faxxled textor cavity, in this regard, provide the results of calculations that Show the impact of ex veuel FCis in a flooded cavity on containment performance. In particular, denumstrate that the effects of an FCI or a rapid steam generation event do not damage equipnwnt or structures rtyuired to nutintain panive containment cooling. As part of this a6wasnent, the potential for and conscquences of ex venel FCis for ludh low pressure and high preuure reactor venel failute 6hould be mnsidered.

Response

As a part of the AP600 PRA, an evale.ation was carried out to examine the potential for and consequences from fuel-coolant interactions, it is dm urnented in the Westinghouse doeunwnt WCAP-13388, " AP600 Phenomenological Evaluation Summaries

• in a section entitled 'A Phenomenological Esaluation Summary on Steam Explosions in Support of the AP600 Risk Analysis.' Apart from the containment shell, all the equipment and structures required to maintain panive containment cooling are outside the containtnent and, therefore, they are not exposed to the fuel-coolant interaction effects. Ilamt on the evaluation described in WCAP 13388, it is concluded that loss of containnwnt function due to a fuel coolant interaction event is not crnlible for the AP600 design.

PRA Revision: NONE

[ WCStingh00S8

NRC REQUEST FOR ADDITIONAL INFORMATION R

Ouestion 720.39 Provide an assessment of the axial and radial ablation of concrete that would occur in the event that water is not added to the reactor cavity.

Response

"Ihe response of the AP600 containment design to stial and radial ablation of the concrete for the highly unlikely use of a dry reactor cavity is addresned in the Westinghouse document WCAP-13388, 'AP600 Phenomenological Evaluation Summaries' in a section entitled 'Phenomenological Evaluation Summary on Molten Core-Concrete Interaction in Suppo:1 of the AP600 Risk Analysis,* The assessment concludes that concrete ablation (MCCI) can be excluded as a credible containment failure mechanism due to the long time (days) required for a sustained dry condition to brewh the AP600 containment, Details of the nuessment are presented in WCAP-13388.

PRA Revision: NONE

[ WC5tiligl10U50

1 l

NRC REQUEST FOR ADDITIONALINFORMATION Ouestion 720.45 The fission prmluct release fractions provided in Chapter 11 of the PRA appear inappropriate for accidents which bypes containment and do not benefit from fiuion product holdup and retention in the containment, in this regard, en additional release class (es) should be added to cover: (1) LOCAs outside containment, (2) interfacing system LOCAs, and (3) steam generator tube ruptures. Provide estinutes of source terna for these 3 additional containment bypass sequences, and an anessment of ther,e source terms on the AP600 risk profile.

Response

The nujor goal of the PRA in to denwnstrate that the AP600 meets the design criterion of limiting the mean whole txxty offaite dose at the site txmndary to less than 25 rem at a frequency greater than 1.0E-6. The release categories were determined with this criterion in mind. Containtnent ivilation failure, and containment bypass (SGTR and ISt.OCA) both result in large source ternu with significant consequences exceeding the dose criterion, and therefore were lumped together. The containment isolation release was used for the dose calculation since it represented the most likely release in the Cl release category at 47% of the release category frequency. Steam generator tube rupture represents 20% of the release category frequency. The remaining 33% is excessive leakage (non-containment failure) sequences resulting in greater than I rem at the site boundary. LOCA outside containment and ISLOCA are not dominant initiating events in the AP600 PRA (see Table 71 of PRA report) since the interfacing systems which extend outside containment have an additional valve and are designed to better withstand the pressuritation expected from multiple valve failures. 'Ihe most likely SGTR sequences result in no releases to the environment (AP600 PRA report section L.2.6) but are classified as core darnage events due to a conservative success criterion uwd in the SOTR core damage frequency calculation. A SGTR sensitivity case which results in large releases is analyin! and discussal in t.cction L.3.7 of the PRA report. The excenive leakage release are much smaller than relene from a containment isolation failure as they result in releases greater than I rem whole txxly at the site boundary. Therefore, the representative release in Cl is appropriate, and significantly overestimates the releases for over half the sequences in the relea.se category. No additional fission prmluct source term estimates are requiral.

PRA Revision: NONE U

W Westinghouse

NRC REQUEST FOR ADDITIONALINFORMATION I

b Ouestion 720.46 To better understand the transport and retention of fiu. ion pnxtucts in the M AAP calculations, provide a break down of the distribution of fission products within the RCS, containment, environment, etc. for each release class in a manner similar to reported in Table 5.8 of NUREG/CR-4624, Volume 1.

Response

The tables are attached in the following pages, and will be included in the PRA report in Chapter 11.

The last paragraph of the following Subations will be revised as follows:

PRA Revision:

(Subsection 11.2.1)

Figures Il 1 through Il 12 present the fission pn> duct release fractions as functions of time for release category OK. Because of the influence of water in the containment, there is cuentially no difference in fission product release if the debris remains in the venel or is released to the containment. The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table 11 1. 'the distribution of the fission products throughout the containment in case BCl is summarind in Table 112.

(Subsection i1.2.2)

Figures 11-13 through 11-24 show the fission product release fractions as functions of time for release category OKP. Because of the influence of water in the containment, there is essentially no difference in fission product release if the debris remains in the vessel or is released to the containment. The final releases, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presenkd in Table 11 1. The distribution of the fission products throughout the containment in caw OKP in summarind in Table 113.

(Subsection i1.2.3)

Figures Il-25 through Il 36 show the fision product release fractions as functions of time for the CC relene category. The final release iractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table 11 1. The distribution of the fission products throughout the containment in case CC is summarized in Table 11-4.'

(Subsection 11.2.4)

Figures 1137 through 11-48 show the fission product release fractions as functions of time for the Cl relcee category. The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presenkd in Table !1 1. The distribution of the fi6sion pnxlucts throughout the containment in cee Cl is summarized in Table 115.

3 WB511righ00S8

NRC REQUEST FOR ADDITIONALINFORMATION Table 112 Distribution of Hwion Products Ily Group - Helcaw Category OK Upper I;2wer Middle Specie- RCS Debris SO Rmuns Compt Compt Annulus Environ ment Xe, Kr 2.8 E-3 0.0 4.lE-2 8.4E l 1. l E-1 9.2E-4 4.21!-5 Cal 3.8E 1 9.1 E-4 5.5E 2 3.8E-1 1.7 E-1 1.4 E-5 5.6E 7 Sr0 1.5E-2 9.7E-l 1.2 E-3 8.BE-3 4.7E-3 7.5E-7 3.2E 8 moo 2 2.6E-l 4.6E 1 2.5E 2 1.8E 1 9.4 E-2 1.3 E-5 5.6E 7 C60ll 3.7E-l 5.7E-4 5.5E 2 3.9E-1 1.7E 1 1.6E-5 5.8 E-7 Ila0 1.4 E-1 7.3E 1 1.2E 2 8 . 1 11- 2 4.4E 2 6.9E-6 2.9E 7 143 0 8.0E 3 9.8E-1 7. 8 E-4 5.5E-3 3.0E-3 4.8E 7 2.0E-8 CeO, 1.90 2 9.6E-1 2.1E 3 1.5E 2 8.2E-3 1.4 E-6 5.9E 8 Sb 1.7 E-l 2.6E 1 3.7E-2 4.0E-1 1.3 E-1 3.7E-5 1.0E-6 720.46 2 W Westinghouse i

NRC REQUEST FOR ADDITIONAL INFORMATION Table 113 Distribution of Fiulon Products fly Group - Rdraw Category OKP Upper lower Middle Species RCS Debris 50 Rooma Compt Compt Annulus Environment Xe,Kr 2.9E = 3 0.0 4.2E-2 8.4E 1 1.1 E-4 9.6E 4 1. l E-4 Csl 4.5E-1 1.2E-3 4.6E-2 3.4E 1 1.5E 1 3.6E-5 2.0E-6 Sr0 3.5E-2 9.5E 1 1.3E 3 8.7E-3 4.7E 3 1.0E 6 8.0E-8 moo, 4.6E 1 2.8 E-l 2.3E 2 1.5 E-1 8.lE 2 1.3E 5 9.6E 7 CxOli 4.3E 1 1.2 E-3 4.8 E-2 3.6E 1 1.6E 1 3.88-5 2.0E-6 Da0 2.8 E-1 5.9E 1 1.2 E-2 7.6E-2 4.1E 2 8.4 E-6 6.5E-7 12 0, 3

2.7E-2 9.7 E-1 6.4 E-1 4.5 E-3 2.4E-3 6.9 E-7 5.5E 8 Ce0 3 6. 8 E-2 9. l E l 2.2E-3 1.5E 2 8.0E-3 2.0E4 1.6E 7 Sb 2.7E 2 1.7E-1 5. 8 E-2 5.1 E 1 2.2E 1 8.8E 5 4,8E4 W westingnouse

NRC REQUEST FOR ADDITIONALINFORMATION Table 114 Distribution of Fiulon Products Ily Group - Heltase Cah1:ory CC Upper lower Middle Species RCS Debrin 50 Moorm Compt Cornpt Annulus Environme nt Xe, Kr 2.81! 3 0.0 4.2E 2 8.3E 1 1.1E 1 9.6E-4 6.5E 5 Col 5.011 1 0.0 4.9E 2 3.0E-l 1.5E-l 1.5E 5 7 . 9 12 - 7 Sr0 2.28 2 9.5 t!-l 1.5f! 3 9.4U 3 5.2E 3 7.6E 7 4.9E 8 Moo, 3.6E 1 3.9 E- 1 2.5 E-2 1.Sfi 1 8.4E 2 1.0E-5 6.5 E-7 CwOli 4.6E l 0.0 5.111 2 3.2E-1 1.6 E-1 2.0E 5 9.0E-7 lho 2.0E-l 6.6E l 1.3E 2 8.2E-2 4.6E 2 6.5 E-6 4.2E 7 1.a203 1.6E 2 9 . 7 11 - 1 9 . 0 13 - 4 5.7E 3 3.2E-3 4.8 E-7 3.lE 8 CeO, 3.7E 2 8 . 9 11- 1 3 . 2 11 - 3 2.0E-2 1. l E-2 1.6E-6 1.18-7 Sb 3.711 1 1.4E 1 3.8E-2 2.5E-1 1.3E 1 2.2E 5 1. l E-6 20m W Westinghouse L-

NRC REQUEST FOR ADDITIONAL INFORMATION Table 115 Distribution of thion l'rodtnts lly Group - Heleaw Category Cl Upper Lower Middle Species RCS Debris $0 Roorm Cornpt Compt Annuha Environment Xe, Kr 9.0E 2 0.0 9.7i3-3 5.lE-l 4.48 2 1.4E-5 3.4E 1 Cal 1.9E-l 1.6E 3 5.7E 1 2.0E 1 6.3E-3 5.6E-7 3.7E-2

$rO 5.lE 3 9.6E 1 3.7E-2 1. l E 3 7.8E 5 1.3 E-9 6.711 5 moo 2 6. l E-2 4.5111 4.7E-1 1.7E 2 8.0E-4 1.9E-8 1.5E 3 CsOli 1.8 E-1 1.6E 3 5.8E-1 2.0E-1 5.lE 3 4.lE 7 3.7E-2 Ba0 3.7E 2 6.BE-1 2.7E 1 7.3E 3 5. l E-4 9.2 E-9 4.8 E-4 1.a,0 3 2.0E-3 9.8E 1 1.3E 2 3.0E-4 2.6E-5 4.BE-10 2.0E 5 Ce0 3 3.0E-3 9.7E 1 2.3E 2 4.7E-4 3.7E 5 6.9E 10 2.8E 5 Sh 4.7E-2 3.3E 1 6.lE l 6.3 E-3 1.4 E-4 9.7E-10 1. l E-3

[ WO5tiflgh0llSe

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 720.47 For the Cl tricaw category, provide a time dependent radionuclide distribution for the containment volumes and primary system. Show the decontamination factors for primary coolant system and containment. Indicate the extent to which the various deposition nnhanisms tontribute to decontamination. In particular, explain why are the zenon and trypton relece fractions less than 100% for a faikd containment, ne 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> source term calculation for this accident newis to be justified.

Response

He time dependent distribution of the fission pn> ducts is includal as Tables 720.47-1 through 720.47-4, each for different times in the analysis of case LFW, which was the representative saluence for release category Cl. The four times in the table relate to: 4 llours - af ter core damage and before the temperature induced hot leg failure, 8 Ilours - after the temperature induced hot leg failure, 20 llours . long-tenn relenes, 28 Ilours - end of the analysis (~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initial core darnage) ne tables show the fraction of the noble gas, volatile and non volatile fission pnx!ucts in the core debris (core region and lower head of the reactor vessel), the RCS piping, the various compartments of the containment, and environment. In the tables, noble gases are represented by the Xe, Kr finion product group, volatile aerosols are represented by the Cal fiuion pnxiuct group and the non-volatile aerosols are reprewnted by the Sr0 fiuion product group. Also included in the tables are the decontamination factors of the RCS and the containment, ne individual deposition rates for the valous aerosol removal mahanisms are not available from the M AAp 4.0 output, but the overall deposition rates are available, expressed as aerosol depositionlambdas. For the ifW case, the lambdas are:

at hot leg failure A = 3.0 hri after hot leg failure A = 0.5 hr' The reason that the noble gas relcue is less than 100% for the Cl release category in that the containment pressure is very near to atmospheric prenure for rnost of the accident time after core damage, and therefore, the driving force for containment leakage in low. Most of the release occurs as the hot leg fails due to high temperature and pressure creep rupture, fission product are released to the containment, and simultaneously the containment is pressurized. After the preuure subsides, rnuch of the decay heat goes into heating RCS metal and into the cavity water which is subcooled water injected from the IRWST. Derefore, long-term containment pressurization is slow, and the rate of the release of fission products to the environment after the failure of the hot leg is slow.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release is justified since the goal of the PRA is to demonstrate that mean whole body offsite doses greater than I rem at the site boundary h:ve a frequency less than 1.0 x 10' per reactor year. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after cose damage, the containment isolation failure cases have significantly exceeded a 1 rem offsite dose, and therefore, the fission product hource term analysis no longer needs to be continued, llowever, the AP600 meets the above criterion as the frequency of the Cl release category is 3.0 x 10' per reactor-year.

PRA Revision: NONE W Westinghouse

'l NRC REQUEST FOR ADDITIONAL.lNFORMATION l

l Table 720.471 l.

l-Fiwlon Product Distribution Release Category CI i Time = 4 Ilours, IYior to llot 1Att Failure RCS SO Imwer Upper Mid RCS Cont =

Species Debris Piping Rooms Compt Compt Annulus Enviro DP DF-Nobles 8.6E 1 1.18 1 8.00 4 2.0E-3 2.4 E 3 8.8E-8 1.2B 3 5.0 23.5 I

Volatile 9.1E 1 8.6E-2 4.3E 5 8.3E 5 9 7E-4 3.0E-9 4.lE 5 76.6 27.7 Non- 1.0 0.0 0.0 0.0 0.0 0.0 ' O0 - -

!' volatile Table 720.47 2 11uion Product Distribution . Release Catqtory CI- ,

Time = 8 liours, After llot Idet Failure ,

l _'

RCS- So lower  : Upper Mid RCS' - Cont -

Species Debris Piping' Rooms Compt Compt Annulus - Enviro DF -DF

- Nobles 2.9E-1 4.9B-4 1.0E-2 3.5B 2. 5,1E-1 3.8E-6 1.6E 1.0 ' 4.4 .

. Volatile . 2.6E 1 3.0E 1 2.1E 1 5.5B 3 1.9E-1 5.5E 7 - 3.6E 2 1.7 -12.2 Non- 9.9E 1 4.55-4 4.95 3 3.2E-6 2.2E-4 9.6B-11' ' l.7B-5 1.1 310 volatile -

t

l. ,

'[

720.u 2 W westinghouse.

= - - -

NRC REQUEST FOR ADDITIONALINFORMATION in l ;4ll Table 720.47 3 Fission l'ruduct Distribution - Release Catq:ory Cl Time = 20 llours,12mg Term Releaw RCS SG 1>>wer Upper Mid RCS Cont Specien Debris Piping Romns Compt Compt Annulus Envito DP DF Nobles 9.0ll-2 0.0 1.1E 2 5.0E-2 5.6E-1 1.lE 5 2.8E-1 1.0 3.2 Volatile 1.611 3 2.0E-1 5.6E l 6.2E-3 2.0E-1 5.6E-7 3,7 E-2 1.3 21.5 Non- 9.6E-1 4.5E-3 3.2E 2 4.6E 5 8 . 2 11- 4 8. l E-10 4.8E 5 1.1 675 volatile Table 720.47-4 1% ion l'roduct Distribution - Release Category CI Time = 28 Ilours,24 Ilours After Core Damage RCS S0 12iwer Upper Mid RCS Cont Specica Debris Piping Rooms Compt Compt Annulus Enviro DP DF Nobles 9.0E 2 0.0 9.7E-3 4.4E 2 5.lE-3 1.4 E-5 3.4E-1 1.0 2.6 Volatile 1.60-3 1.9E-1 5.7E-1 6.2E 3 2.0E 1 5.6E 7 3,7 E-2 1.2 21.9 Non. 9.6E 1 5. l E-3 3.7E-2 7.88-5 1. l E-3 1.3 E-9 6.7E-5 1.1 556 volatile W Westinghouse i

i 1

NRC REQUEST FOR ADDITIONAL. lNFORMATION III" i I

!F 1

. e Question 720.50 On page 10-1 of the PRA, it is stated that NUREG-1335 was um! as a guide for seixting model parameters to be varied in the sensitivity studies. Clarify whether the EPRI guidance document on MAAP sensitivity analyses for the IPE was also used, and why it was not, if that is the case.

Response

The EPRI guidance docunent ' Recommendations on the Use of the MAAP 3.0D Code in Individual Plant Evaluations' was used to a certain extent in the selection of the e2600 PRA sensitivity studies, llowever, the PRA was performed with M AAP 4.0, not M AAP 3.00, and the AP600 does not always respond in the same manner an a conventional plant, for which the guidance was written. For example, the discussions of sensitivity due to the M AAP 3.0B core blocLage trnlel do not apply to M A AP 4.0, since tl.ere is no core blockage rnodel in M AAP 4.0.

Similarly, the monsitivity to containnwnt failure parameters do not apply to AP603, since the containment integrity is not threatened in sequences in which the containnent is not failed prior to, or at the initiation of an accident.

Analyses to addreas the severe accident issues recommended in the guidance doeurnent have been performed in the PRA. Specifically, the t.ensitivities of M AAP 4.0 parameters which affect the rate, or occurrence of core-concrete interaction, in-venel hydrogen generation, and hydrogen combustion were perfonned. Supporting analyses were performed to addreas timing and location of high pressure and temperature induced failures of reactor coolant system piping. He above mentioned and other severe accident issues, such as high pressure melt ejection, fuel coolant interactions, and fission pnxluct deposition are addressed in WCAP-13388, 'AP600 Phenomenological Evaluation Summarica". In addition, in the WCAP, external cooling of the scactor vessel lower head is also addressed, and not specifically recommended in the EPRI guidance document.

PRA Revision: NONE i

720.50 1 WO5tingh0USB

l NitC ftEQUEST FOlt ADDITIONAL. lNFORMAllON Question 720.57 Confirm that the June 26,1992 l'R A reflats all of the thanges inade to the AP600 design, as presented in the SS AR submitted on June 26, 1992, or identify the dif ferences between the dnign muunni in the PRA and that of the design application.

flesponse.

There are casen where dillerences exist between the design assumed in the PR A model and the design dewribed in the AP600 SSAR. Thet.e diflereneca, and proje(tal effc(ti, are identified in Appendicies C8, C9, Cil, Cl2, Cl3, Cl5, C16 Cl7, C21, and C22 of the PRA report.

PRA flovision: NONE

[ WC51lDgt100$0

NRC REQUEST FOR ADDITIONAL. lNFORMATION Ouestion 720.58 To perform confirmatory analyses of the Westinghouse AP600 PRA results, the staff is planning to upload the Wentinghouse AP600 PRA onto the IRRAS wmputer program. To upload the Westinghouse PRA onto IRRAS, the staff neals to have the following files on electronic malia in ASCil fornat, unless otherwise stated. Provide this information.

a. For all of the fault treca generated from Grafter, the staff necds to have all of the treename. tat and troename.dat files converted to SETS input using the SETSIN2 program as deacribal in the ORAITER Users Manual (WCAP ll693). The subtreen designated as bull-XXXX also rmx! to be convertal to SETS input. The staff a!*o needs a copy of all of the fault tree output files frorn ORAITER (treename. cut), including the SUll XXXX subtrees.

b. The staff needs the Master Data file from OR AITER umi to quantify the bad: events that are describal in the fault trecs. The staff believes that the file is called SIMON.DAT or SIMON. CUT.

c. Ilami on conversations with Westinghouse, the ORAITER computer output given to the staff contains incomplete system cutsets and incomplete system unavailabilities. 'these systems may contain basic events ,

that are designated as SUD XXXX that repreacnt smaller subtrees that are given dummy probabilities. The staff understands that these system fault treen are reduced (the SUD XXXX basic events are replaced with cutsets) in the SUllA option in the WLINK computer code. Therefore, the staff needs the fault trees output after the SUBA option is uml in WLINK that reduces all of the SUil XXXX events. The staff believes that these are treename.wik files. The stalf also needs to have a copy of the accident anguence output files from the SEQ OPTION in WLINK. The staff believea that they are called XXXX.out files.

Responso;

'Ihe requested files were provided on diskettes in letter ET-NRC-92-3774 on Novemh:r 25, 1992.

FRA Revision: NONE 8

W Westinghouse