ML20237B512

Forwards Request for Addl Info Re Pda Modules 1-16 of RESAR SP/90

ML20237B512
Person / Time
Site:	05000601
Issue date:	12/09/1987
From:	Kenyon T Office of Nuclear Reactor Regulation
To:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 8712160317
Download: ML20237B512 (21)

Text

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, . December 09, 1987 Docket No. 50-601 Mr. W. J. Johnson Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division Box 355 Pittsburgh, Pennsylvania 15230

SUBJECT:

REQUEST FOR ADDITIONAL INF0PMATION ON RESAR SP/90 As a result of our ongoing review of PDA Modules 1 through 16 of the RESAR SP/90 application, we require additional information in order to complete our review of the mechanical engineering aspects of the design. Enclosed are review questions Q 210.25-210.72.

Please respond to this request within 90 days of the date of this letter. If you have any questions regarding this matter, call me at (301) 492-8206.

Sincerely, original signed by Thomas J. Kenyon, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page I

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Docket No. 50-601 Mr. W. J. Johnson Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division Box 355 -

Pittsburgh, Pennsylvania 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 As a result of our ongoing review of PDA Modules I through 16 of the RESAR SP/90 application, we require additional infonnation in order to complete our review of the mechanical engineering aspects of the design. Enclosed are review questions Q 210.25-210.72.

Please respond to this request within 90 days of the date of this letter. If you have any questions regarding this matter, call me at (301) 492-8206.

Sincerely, l

Tho s J. Ke n, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As. stated cc: See next page

,, Docket No. STN 50-601 RESAR-SP/90  :

CC:

Trevor Pratt Brookhaven National Laboratory Building 130 Upton, New York 11973 Mr. William Schivley Westinghouse Electric Corporation ECE-410 Mail Stop 4-08 Box 355 Pittsburgh, Pennsylvania 15230 l

i

'V REQUEST FOR ADDITIONAL'INFORMATION RESAR SP/90 - WESTINGHOUSE ADVANCED PRESSURIZED WATER REACTOR - DOCKET NO. 50-601 HECHANICAL ENGINEERING BRANCH I. MODULE 1 - PRIMARY SIDE SAFEGUARDS SYSTEM MODULE l

1 210.25 The information in Table 1.8-2, Section 3.2.2. and Section 3.2.3 l

relative to the WAPWR alternatives to Regulatory Guide 1.26 is j not currently acceptable. Specifically, the staff has not en-dorsed the detailed guidance in ANSI /ANS 51.1 - 1903 to deter-mine the quality group classification of systems, components and equipment which ere important to safety as defined in the Introduction to 10CFR 50, Appendix A. A discussion of the staff position on this issue is contained in question 210.35 on Module

7. Subsequent to a resolution of this issue, the inf6rnation on Reg. Guide 1.26 in Table 1.8-2 Section 3.2.2 and Section 3.2.3 of Module 1 should be revised to agree with the response to Q210.35.

II. MODULE 2 - REGULATORY CONFORMANCE 210.26 Section6.5.2ofModule2referencesSection6.5.10asaEAPWR response to IE Bulletin 79-02, " Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts." Section 6.5.10 l addresses IE Bulletin 80-11. "Masonary Wall Desigr." The staff does not agree that IE Bulletin 80-11 is a sequel to IE Bulle-tin 79-02 for anchor bolts. The information requested in 80-11 does not adequately address the information requested in 79-02.

l Revise Section 6.5.2 of Module 2 to provide sufficient information to respond to IE Bulletin 79-02.

III. MODULE 4 REACTOR COOL %NT SYSTEM 210.27 The staffs' comments in 0210.25 and 210.35 also apply to

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portions of Table 1.8-2, Section 3.2.2, Section 3.2.3 and Section 5.2.2.6 of Module 4. These sections should be revised to agree with the response to 0210.35. 1 I

210.28 The information in Table 1.8-2 which discusses exceptions taken  !

to Regulatory Guides 1.124, Revision 1 and 1.130 Revision 1 does not completely conform the current staff positions rela-tive to design criteria for /SME Class 1 component supports.

To be. acceptable, this information should be revised to provide a commitment to construct all Class I component supports in accordance with the rules rf AShE III, Subsection NF, "Compo-nent Supports", 1986 Edition or to the Code of Record which will be applicable to the final WAPWR plant. (Reference Questions 210.60 and 210.65)

IV. MODULE 5 - REACTOR SYSTEM 210.29 The staffs' comments in 0210.25 also apply to portions of Section 3.2 and Table 3.2-1 of Module 5. These sections should be revised to agree with the response to Q210.35.

210.30 Section 1.5.1.3 of liodule 5 briefly discusses proposed tests of the Control Rod Drive Mechanism (CRDM) and the Water Displacer Rod Drive Mechanism. These tests were scheduled to be conducted in 1!85 and 1086. Provide a detailed description of the test program and a summary of the results. If applic-able, provide a comparison of tests which were conducted on existing Westinghouse CRDM's with the WAPWR CRDM's.

210.31 In Section 3.9 2.4 of Module 5, it is stated that the recom-mendations of Regulatory Guide 1.20 will be satisfied by con-l ducting examinations of the reactor internals both before and after confirr.atory hot functional testing of the internals.

3 Based on the staff's understanding of the WAPWR reactor inter-nals design, this is not an acceptable cornitment. Section 3.9.5 of Module 5 describes a design which is " significantly different from existing Westinghouse designs". The prototype plant for existing Westinghouse four loop plants is Indian Point Ur.it 2. The data from the Indian Point 2 reactor internals verification test prnpram has been supplemented by data from tests conducted at the Trojan and Sequoyah plants. This supplemental data was provided at the staff's reauest to verify that design changes to reactor internals in Westinghouse four loop plants subsequent to the Indian Point 2 design did not result in a significant difference from the Indian Point 2 verification data.

It is not apparent to the staff that the WAPWR reactor inter-nals response to flow induced vibration is enveloped b'y the above prototype verification data for four loop plants. There-fore, the staff will require a commitment that the first WAPWR plant will be identified as the prototype plant and will meet all of the applicable Regulatory Guide 1.20 guidelines. Revise Section 3.9.2.4 in Module 5 to provide this commitment, or provide justification for not doing so.

210.32 As stated in Q210.31, the staff position is that the first WAPER plant be designated as the prototype as defined in REG Guide 1.20. The information in Section 3.9.2.3 of Module 5 relative to the dynamic response analysis of reactor internals under operational flow transients and steady-state conditions

, is not acceptable for a prototype plant. Revise this section to be consistent with the guidelines in Stendard Review Plan, Sections 3.9.2.I.3, 3.9.2.11.3 and 3.9.2.111.3.

210.33 Section 3.9.5.1.3.4. of Module 5 discusses the botton mounted instrumentation (BMI) thimbles. A problem of unacceptable accelerated wear of Westinghouse designed BMI thimbles in a European 14-foot core plant wes identified in 1985.

4 Subsequently, Westinghouse modified the thimble design to re-duce flow velocity in the gap between the thimble and the BMI column. However, this modification did not resolve the problem, but instead increased the rate of wear. The same modified design has been incorporated into the South Texas, Unit 2 pressure vessel which also contains a 14-foot core. Since this potential problem could be applicable to the VAPWR, provide a commitment in Section 3.9.5.1.3.4 that the final resolution of this problem for South Texas and the European plants will be incorporated into the WAPWR design, or provide justification for not doing so. A failure of one or more of these thimbles could result in a small break in the reactor coolant pressure boundary which cannot be isolated.

'i V. MODULE 6 - SECONDARY SIDE SAFEGUARDS SYSTEM 210.34 The staff's comments in Q210.25 and 210.35 also apply to portions of Section 3.2.2. and 3.2.3 of Module 6. These sec-tions should be revised to agree with the response to 0210.35.

VI. MODULE 7 - STRUCTURAL / EQUIPMENT DESIGN I

210.35 The responses to questions 210.1, 210.3, 210.6, 210.7, 210.8 210.9. 210.12, 210.13, and 210.16, are not acceptable. The classifications in Table 3.2-i of Module 7 in the Quality Group, Safety Class Seismic Category and Q-List columns in all of these responses are not consistent with the guidelines of Regulatory Guides 1.26 and 1.29. As stated in question 210.1, the staff has not endorsed ANSI /ANS 51.1-1983 and cannot use

, this document in determining the acceptability of the classifi-cation of structures, systems and conponents. Provide a revi-sion to Table 3.2-1 in Module 7 to be consistent with the above l staff position, or provide further justification for non-l compliance. In addition, the ASME Section III Code Class should 1

I correspond to the guidance provided in REG Guide 1.26 i

5 210.36 The responses to questions 210.2, 210.18 and 210.22 are not acceptable. In the staffs' opinion, Footnote S to Table 3.2-1 in Module 7 places too much emphasis on the judgement of indi-  !

viduals to make a decision on the classifications of a support.

This could result in an inconsistent and potentially uncenserva- ,

tive approach to such classifications. As stated in question 210.2, the staff position on this issue is that supports and hangers should be classified to the same Quality Group ard Safety Class as the component that is being supported. Revise Table 3.2-1 in Module 7 to modify or delete Footrote S to be consistent with the above position, er provide justification for not doing so.

210.37 The Westinghouse Positions 2(B) and 3 in the response to question 210.19 are not acceptable. Plant specific safety-related instrument tubine programs which did not completely conform to Regulatory Guide 1.151 have been accepted by the staff for several recently licensed plants. The staff's basis for approving these programs was that the plants had been designed long before the issuance of Reg. Guide 1.151. For future plants, the staff requires that safety-related instrument sensing lines be designed in accordance with Reg. Guide 1.151.

Revise the response to Q210.19 to provide a commitment to this staff position, or provide justification for not doing so.

210.38 The response to question 210.21 is not completely acceptable.

All of the plants which have been licensed by MRC so far have been allowed to request relief from the ASME Section XI inservice testing rules for a limited number of pumps and valves. These pumps and valves are generally installed in systems in which it is impractical to meet the Section XI rules because of limitations in the system design which make the pump or valve difficult to test without additional design changes.

Therefore, the staff granted many of these reouests for relief

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because imposition of these rules would have resulted in hard-ships to the licensee without a compensating increase in the level of safety. The underlying reason for the regulation allowing these reliefs from the code was that the detailed piping systen designs for all of these plants was completed prior to the time that the staff began to implement the ASME Section XI rules.

A plant such as the WAPWR, for which the final design is not complete, has sufficient lead time available to include pro-visions for this type of testing in the detailed design of applicable piping systems. Therefore, requests for relief from the applicable ASME Section XI testing rules for pumps and valves will not be granted for the WAPWR. Revise the response to 0210.21 to provide a commitment that PAPWR piping systems will be designed to accommodate the applicable code re'quirements for inservice testing of pumps and valves. However, with regard to subsequent or future code revisions to the applicable ASME Code for WAPWR, requests for relief from certain updated code requirements may still be submitted for staff review in accordance with 10CFR50.55a(g).

210.39 The information in Sections 3.6.2.1.1.A. 3.6.2.1.1.B and 3.6.2.1.1.C of Module 7 relative to postulation of pipe ruptures in high energy ASME Class 1, 2, and 3 piping and non-nuclear piping is not consistent with current staff positions on this subject. These positions are in Standard Review Plan, Section 3.6.2, Branch Technical Position MEB 3-1, Revision 2 dated June, 1987. Revise these three sections in Module 7 to be consistent with MEB 3-1, Revision 2, or provide justificetionfor not doing so. This revision should not only include changes in threshold stress and curulative usage factor levels, but should add all of the other MEB 3-1 guidelines which are currently not included in l

I the WAPWR high energy break location criteria.

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210.40 The information in Section 3.6.2.1.1.D in Module 7 relative to high energy ASME Class 2 piping in containment penetration (break exclusion)' areas is not completely acceptable. In addi-tion to the unacceptable stress levels discussed in 0210.39, the following revisions to this section are required to cualify for break exclusion in penetration areas:

1. A 100% volumetric inservice examination of all pipe welds should be conducted during each inspection interval

,as defined in IWA-2400 ASME Code,Section XI.

2. If there is any ASME Class 1 high energy piping in contair. ment penetration areas in which breaks are not postulated, provide the basis for not postulating breaks in these systems.

Revise Section 3.6.2.1.1.D in Module 7 to include the above information, or provide justification for not doing so.

210.41 The information in Section 3.6.2.1.1.E in Module 7 relative to

" Leak-Before-Break" criteria needs to be updated. The current staff position on this issue is contained in the Draft Standard Review Plan 3.6.3, " Leak-Before-Break Evaluation Procedures" dated August, 1987. This draft was recently sent out for pub-lic comments in the Federal Reaister Notice, Vol. 52, No. 167,

p. 32626-32633, dated August 28, 1987. This document contains complete and up to date guidelines which the staff will use to determine the acceptability of plant-specific Leak-Before-Break submittals.

Revise Section 3.6.2.1.1.E in Module 7 va be consistent with the guidelines in the above document.

8 210.42 Section 3.6.2.1.2, " Types of Breaks / Cracks Postulated" in Module 7 does not discuss criteria for postulating cracks in j high-energy piping. Revise this section to provide this infor-

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mation. Acceptable guidelines can be found in Standard Review

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Plan 3.6.2, MEB 3-1 Section B.1.e. Revision 2 dated June 1987. {

Piping covered by Sections 3.6.2.1.1.0 and 3.6.2.1.1.E in )

Module 7 are excluded from these guidelines.

210.43 In Section 3.6.2.1.2.3 of liodule 7, the threshold stress values for postulating cracks in moderate energy piping does not agree with the current staff position which is in SRP 3.6.2.

Specifically, in Sections 3.6.2.1.2.3.A and 3.6.2.1.2,.3.C.(2), ,

0.45 should be 0.40 and in Section 3.6.2.1.2.3.C.(1), 1.4 Sm should be 1.2 Sm. Revise this section to be consistent with the staff position, or provide justification for deviation from the guidelines ir,SRP 3.6.2.

210.44 Sections 3.6.2.1.2.1.A.1 and 3.6.2.1.2.1.A.2 in Module 7 contain an apparent typographical error. To be consistent with the staff positen on this issue, "circumferential break" should

. be " longitudinal break" in paragraph 1 and " longitudinal break" should be "circumferential break" in paragraph 2.

210.45 Section 3.6.2.2.1 in Module 7 references WCAP 10221,

" Simplified Pipe Whip Analysis and Restraint Design Procedures" for analytical methods used in calculating jet thrust loads subsequent to a pipe rupture. The staff has no record of re-ceiving this report and has not reviewed it. Standard Review Plan 3.6.2 and ANS 58.2, " Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Breaks" contain acceptable guidelines for these analytical methods. Revise Section 3.0.2.2.1 to either reference the above documents or provide a description of the methodology which will be used to calculate jet thrust loads.

9 210.46 In addition to the information in Section 3.6.2.3.4.1 of Module 7, provide the loads, load combinations, and stress limits thet will be used in the design of pipe rupture re-straints. Include a discussion of the design methods applica-ble to the auxiliary steel used to support the pipe rupture restraint. Provide assurance that the pipe rupture restraint and supporting structure cannot fail during a seismic event.

Provide the design criteria which will be used for pipe rupture restraints that also support piping, if this criteria is dif-ferent from that discussed in Sections 3.9.3.1,1 and 3.9.3.1.2 of Module 7.

210.47 Section 3.7.3.1 of Module 7 references the ASCE Seismic Analysis Standard Committee ' Standard for the Seismic Analysis of Safety-Related Nuclear Structures," May 1984 Draft for meth-odology used in both time-history solutions and response spec-trum analyses of subsystems. This ASCE standard (including the Septerber 1986 Edition) has not been accepted by the staff because some of the basic assumptions and options in the stan-dard do not agree with current staff peritions. Revise Section 3.7.3.1 in Module 7 to be consistent with applicable guidelines in Standard Review Plan, Sections 3.7.1, 3.7.2 and 3.9.2, or provide justification for deviations from these guidelines.

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210.48 Section 3.7.3.2 of Modulo 7 states that the Operating Basis Earthquake (OBE) is assumed to occur five times over the life of the plant. A time history study was conducted by Westing-house to arrive at a realistic number of noximum stress cycles per OBE occurrence. As a result of this study. Westinghouse concluded that 10 maximum stress cycles for flexible equipment (natural frequencies less than 33 Hz) and 5 maximum stress cycles fnr rigid equipment (natural frequencies greater than 33 Hz) for each OBE occurrence should be used for fatigue evalua-tien of WAPWR ASME Class 1 systems and components. However,

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10 Table 3.9-1, " Summary of Reactor Coolant System Design Transi-ents" in Module 7 lists 50 cycles for OBE.

It is the staff's understanding that Westinghouse generally uses only the 50 stress cycles criterion for all ASME Class 1 systems ar.d components, which is consistent with the guidelines in Standard Review Plan, Section 3.9.2.II.2.b. Either delete the reference to the use of 5 stress cycles per event for rigid equipment or provide additional justification for using only 25 cycles instead of 50 cycles in the fatigue analysis for this equipment.

210.49 Section 3.7.3.3.A of Module 7 states that for the analysis of main piping runs, branch connections ere decoupled from the main runs when the ratio of the branch to run section moduli is equal to or less than 1/16, or the ratio of the branch to run moment of inertia is 1/50. It further states that the boundary of each decoupled model contains a sufficiently long region of commor overlap to other models. Provide more information reletive to the basis and justification for each of these assumptions.

In the first paragraph of Section 3.7.3.7, Module 7, an 210.50 optional method of algebraic combination of modes with closely spaced frequencies is stated. NUREG-1061, Vol. 4 " Report of the U.S. NRC Piping Review Committee" is referenced as the basis for this option. It should be noted that NUREG recommendations should not alweys be interpreted as staff positions unless 'an explicit statement to this effect is documented in the NUREG.

In this case, the NUREG-1061 recommendation of algebreic combination has not yet been accepted by the staff. The current staff position as stated in Regulatory Guide 1.92. " Combining Modal Responses and Spatial Components in Seismic Response Analysis" has three options in combining closely spaced modes, l but the proposed algebraic combination is not one of them.

Revise section 3.7.3.7 to delete the proposed optional method, or provide justification for its use.

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11 210.51 Sections 3.7.3.7.A 3.7.3.7.B, 3.7.3.7.C. 3.7.3.7.D and 3.7.3.9.A.2 of Hodule 7 discusses several options which will be availtbleintheseismicsubsystemanalysesoftheFAPWR. The bases for these options are provided in References 10, 11, 12, and 13 of Section 3'7 of Module 7. If Westinghouse intends to use these options, the staff will be required to review the applicable references. Therefore, submit the referenced docu-ments for staff review.

210.52 The last sentence in Section 3.7.3.9.B of Nodule 7 states that the effect of relative seismic anchor displacements are obtained by either using the worst combination of peak displacements or l by proper representation of the relative phasing characteristics l 5

associated with different support inputs. Provide more details relative to how proper representation" is obtained. Identify '

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and justify any deviations from the guidelines in Standard 1 Review Plaa, Section 3.9.2.II.2.g. l 210.53 Combination of inertial responses and seismic anchor movements of multiply-supported items by the SRSS method as stated in Section 3.7.3.9.C of Module 7 is not acceptable. Revise this section to provide a comitment to combine these items by abso-lute sua as recommended in Standard Review Plan, Section 3.9.2.II.2.g, or provide justification for use of this method..

210.54 The information in Section 3.7.3.13 of Module 7 relative to the interaction of other piping systems with Seismic Category 1 piping is too general to be completely acceptable. Revise this section to be consistent with the guidelines in Standard Review Plan, Section 3.9.2.II.2.k. In addition, revise applicable portions of Section 3.2.1.1 of Module 7 to be consistent with the revised Section 3.7.3.13.

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12 210.55 The information in Figure 3.7-8 of Module 7 relative to l 1

damping values is not completely acceptable. Specifically, curves (1) and (2) exceed dampirp values currently acceptable i l to the staff. Curve (3) is consistent with ASME Code Case N-411 and is acceptable for all ASME Class 1. 2, and 3 piping provided a commitment is made to conform to the conditions specified by the staff for using Code Case N-411. These conditions are .(

outlined in Regulatory Guide 1.84, Revision 24, dated June 1986. f Revise Figure 3.7-8 to eliminate curves (1) and (2) and to provide a commitment to the conditions of REG Guide 1.84, Rev.

24 for use of curve (3). In addition, revise the portion of the discussion in Sections 3.9.3.1.1.1.C and 3.9.3.1.3 of Module 7 which is applicable to this issue.

I 210.56 The information in Section 3.9.1.2.1 of Module 7 relative to computer codes which will be used in dynamic and static analys-es of seismic Category 1 components and equipment is not con-plete. Provide e discussion in this section of the methods used to verify the FATCON and WESAN programs and any other applicable -

program which is not listed. This verification should be in accordance with the guidelines in Standard Review Plan, Section 3.9.1, i.e., a comparison of the results from each pregram with either (1) hand calculations (2) published analytical results, (3) acceptable experimental results, (4) results from a similar program which has been accepted by the staff, or (5) the bench-mark problems in NUREG/CR-1677, " Piping Benchmark Problems."

The WECAN program, which is also listed in Section 3.9.1.2.1, has received only a partial review by the staff. The informa-tion in WCAP-8929, " Benchmark Problem Solutions Employed for Verification of the WECAN Computer Program" which is applicable to the dynamic analysis of linear and nonlinear elastic beam-type structures is the only portion of WECAN which has been accepted by the staff. However, WECAN contains many other features and extensive capabilities such as plate and shell

13 structures, elastic-plastic and creep deformation and heat )1 l

transfer analysis. None of these features have been independent- )

ly verified, although their theoretical bases are consistent with present state-of-the-art. If any of these additional WECAN featureswillbeusedintheder.ignoftheHAPWR,discussthe application of the feature in Section 3.9.1.2.1 of Module 7 and the staff will review this information on a case-by-case basis. l 210.57 Section 3.9.1.4 " Consideration for the Evaluation of the Faulted Condition" in Module 7 references Section 3.9.4 of j Module 7. Section 3.9.4 only addresses analyses of Control Rod l Drive Systens. The intended reference appears to be Section 3.9.3.4, " Component and Piping Supports." However, Section j 3.9.3.4 does not completely address the subject matter cf Sec- )

tion 3.9.1.4 because it only discusses Service Level D analyses of component supports in the elastic range. If an elastic- ,

plastic method of analysis will be used to evaluate the design of any safety-related system, component er equipment for which ASME Level D Service Limits have been specified, provide the information requested in Standard Review Plan, Section 3.9.1.11.4 and 3.9.1.111.4. This information should be in Section 3.9.1.4 of Module 7.

210.58 The information in Section 3.9.2.1 of Module 7 provides a general discussion of the proposed piping preoperational test program for the WAPWR. The staff's current position on this issue is that for Preliminary Design Approval, a commitment is required to develop a test program for Final Design Approval which will utilire testing procedures and acceptance criteria in the Draft ANSI /ASME OM-3, " Requirements for Preoperational and Initial Start-Up Vibration Testing of Nuclear Power Plant Systems." The staff is participating in the development of this i stancard. BythetimeaEAPWRFinalDesignApprovalissub-mitted, the staff plans to have OH-3 referenced in Standard

14 Review Plan, Section 3.9.2. For the past several years, the staff has been using the acceptance criteria from the OM-3 Draft dated May 1985 as a guide in its review of piping preoperational test prograns for near-term operating license plants. Revise Scetion 3.9.2.1 to provide a commitment to the staff position described above, or provide justification for deviation from this position.

210.59 It is the staff's position that all essential sefety-related instrumentation lines should be included in the vibration moni-toring program during preoperationel or startup testing. He require that either a visual or instrumented inspection (as appropriate) be conducted to identify any excessive vibration that could result in fatigue-failure. Revise Section 3.9.2.1 in Nodule 7 to include a corrnitment to test such piping, or provide justification for deviation from this position'.

210.60 Prov de the following revisions to Tables 3.9-3 and 3.9-5 in Module 7. or provide justification for not doing so:

1. In the Stress Criteria Column of Class 1 Pumps for Service Levels A, B, C and D, add NB-3400. (Table 3.9-3 only)

2. Footnote (a) in Table 3.9-3 states that a test of compo-nents may be perforred in lieu of analysis. Revise this footnote to provide a commitment that if this option is used for the WAPWR, details of the test program will be subnitted to the staff for review prior to implementation of the option.

3. Revise the Stress Criterie Column for component supports in both tables to be consistent with the response to Ques-tions ?10.28 and 210.65, i.e., commit to use ASME Section III, Subsection NF, 1986 Edition.

15 e

210.61 'The staff considers valve discs to be a part of the pressure boundary and as such should have allowable stress limits. If the stress limits in the valve column of Tables 3.9-3 and 3.9-5 Module 7 do not include valve discs provide this criteria in Section 3.9.3.1 of Module 7.

210.62 In addition to the information in Table 3.2-1 of Module 7, revise Section 3.9.3 to include the design basis which will be used to insure ~the structural integrity of. safety-related Heat-

'ing. Ventilation and Air Conditioning (HVAC) ductwork and supports.

210.63 Section 3.9.3.1.3 of Module 7 states that valves in sample line.s connected to the RCS are not considered to be ASME Class 1 because the loss of flow through a severance of one of these 3/8-inch lines can be made up by normal charging flow.' The staff agrees that such piping does not have to be Class 1, however, it must be classified as Ouality Group B and Safety Class 2. On Sheet 19, Table 3.2-1 of Module 7, RCS sample valves and piping meet these guidelines. Verify that these valves and piping in Table 3.2-1 are the only components af-fected by the statement in Section 3.9.3.1.3. If not, provide the classification of other applicable RCS sample valves and piping.

210.64 The information in Section 3.9.3.2 of Module 7 relative to pump and valve operability assurance is not completely accept-able. Sections 3.9.3.2.1.C. 3.9.3.2.2.D, 3.9.3.2.3 and r3.9.3.2.4.C refer to IEEE Standard 344-1975. To be acceptable, these references should also include a commitment to meet the additional guidelines in Regulatory Guide 1.100. Revise all of the above sections and any other applicable section in the PDA to include such a commitment, or provide justification for not doing so.

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210.65 The information in Sections 3.9.3.4.1 and 3.9.3.4.2 of Module l 7 relative to the use of ASME Section III, Subsections NF and Appendix F in the design of ASME Class 1, 2 and 3 component supports needs to be updated. The staff has potential questions on the following subjects relative to Subsection NF:

1. The bases for the selection of NF vs AISC jurisdictional boundaries for supports is requested. Describe which part of the support will be constructed as NF and which part will be constructed as building steel.

2. Provide bucklino criteria used in the design of all component supports.

3. Are the stresses in supports which are produced by seismic anchor point motion of the supported piping and t'he ther-mal expansion of supported piping treated as primary or secondary stresses?

The staff's positions on the three issues above have been in-corporated into the 1986 Edition of Subsection NF. Therefore, a commitment that ASME Section III, Subsection NF, 1986 Edition will be used in the construction of all ASME Class 1, 2 and 3 supports will suffice. It should be understood that "Construc-tion" is an all-inclusive term as defined in ASME Section III, Subsection NB/NC/ND 1100. Revise Sections 3.9.3.4.1 and 3.9.3.4.2 of Module 7 and Section 5.2.27 of Module 2 to provide this corritment. If this commitment cannot be made, provide a detailed response to the above three questions.

The information relative to the test load method in Appendix F should also be revised to be consistent with ASME Section III, Appendix F, 1986 Edition.

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17 210.66 The following additional information is required in Section 3.9.3.4 of Module 7 relative to the design of bolts for compo-nent supports: ,

1. Provide the allowable stress limits which are applicatie to bolts used in ecuipment a;1chorage, component supports and flanged connections. The staff position is that the stress in these bolts should not exceed the yield strength at temperature. f

2. Provide a discussion of the design methods applicable to expansion anchor bolts used in component supports.

210.67 Section 3.9.4 of Module 7 states that in the analyses of the control red drive mechanisms and the gray rod drive mechanisms, a nonlinear elastic LOCA analysis and a separate linear elastic seismic analysis is performed. Provide the basis for perform-ing both a linear and nonlinear analysis on the same structure.

210.68 Section 3.9.6 of Module 7 states that the inservice testing (IST) program will be prepared within 6 months after the WAPWR operating license issue date. This is not an acceptable sched-ule. To provide the staff sufficient time to review the appli-cable information and prepare a Safety Evaluation Report, the IST program must be submitted as a part of the Final Design )

Approval. Revise Section 3.9.6 to provide this commitment.

210.69 The information in Sections 3.9.6.1 and 3.9.6.2 of Module 7 infers that only ASME Class 1, 2 and 3 pumps and valves will be

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includedintheinservicetesting(IST)programforthe}lAPWR. )

It is the staff's position as stated in Standard Review Plan, {

Sections 3.9.6.II.1 and 3.9.6.II.2 that all pumps and valves which are considered as safety-related should be included in  !

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the IST program even if they are not categorized as ASME Class 1, 2 or 3. Revise Sections 3.9.6.1 and 3.9.6.2 of Mcdule 7 to j clarify this commitment. l

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18 210.70 Section 3.9.6.3 of Module 7 discusses relief recuests from the IST requirements of ASME Section XI. Revise this section to be consistent with the response to Q210.38.

210.71 Section 6.5.2.5 of Module 2 contains a brief discussion of the Westinghouse response to the licensing issue relative to periodic leak testing of pressure isolation valves. A more detailed commitment on this issue should be provided in Section 3.9.6 of Module 7. At the Preliminary Design Approval staga, the staff requires a commitment to perform periodic leak testing of all pressure isolation valves in eccordance with applicable requirements in the Westinghouse Standard Technical Specifica-tions. In addition, a commitment is required to provide, at the Final Design Approval stage, a list of applicable valves for the staff's review. Revise Section 3.9.6 of Module 7 to add these commitments.

VII. MODULE 8 - SECONDARY SIDE SAFEGUARDS SYSTEM / STEAM AND POWER CONVERSI0f! SYSTElts 210.72 The staff's comments in Q210.25 and 210.35 also apply to portions of Section 3.2 in Module D. This section should be revised to agree with the response to Q210.35.

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