ML20154Q174

Forwards Request for Addl Info Re Application for Certification of Advanced BWR Design.Responses Requested by 881130

ML20154Q174
Person / Time
Site:	05000605
Issue date:	09/26/1988
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 8810030333
Download: ML20154Q174 (26)

Text

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O O September 26, 1988 Docket No. STN 50'-605 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Operations Mail Code 682 ',

175 Curtner Avenue San Jose, California 95125

Dear Mr. Marriott:

1

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APFLICATION FOR CERTIFICATION OF THE ABWR DESIGN  !

In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.  :

Our request for additional information, contained in the enclosure, addresses  :

the areas of SRP Chapter 3 reviewed by the Mechanical Engineering Branch.

This completes the initial request for additional information on the ABWR -

related to SSAR Chapters 1, 2 & 3. However, the need for additional infor- i mation may occur during the development of the staff's safety evaluation. If this should occur, the need will be identified in a draft Safety Evaluation i 4 Report which will be provided for your consideration.  ;

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by November 1988. If you have any '

concerns regarding this request please call me on (30130,)492-1104.

P Sincerely, Dino Scaletti, Project Manager l 8810030333 000926 Standardization and Non-Power  :

PDR ADOCK 05000605 Reactor Project Directorate A PNV m. Division of Reactor Projects - III. IV,  !

Y and Special Projects  !

Office of Nuclear Reactor Regulation l

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o September 26, 1988 Docket No. STN 50-605 Mr. Patrick W. Harriott, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Operations Mail Code 682 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.

Our request for additional information, contained in the enclosure, addresses the areas of SRP Chapter 3 reviewed by the Mechanical Engineering Branch.

This completes the initial request for additional information on the ABWR related to SSAR Chapters 1, 2 & 3. However, the need for additional infor-mation may occur during the development of the staff's safety evaluation. If this should occur, the need will be identified in a draf t Safety Evaluation '

Report which will be provided for your consideration.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by Novenber 1988.

concerns regarding this request please call me on (30130,)492-1104.If you have any '

Sincerely,

%: Dino C. 3M caletti, Project Manager  !

Statidardization and Non-Power Reactor Project Direct'Jrate Division of Reactor Projects - !!!. IV, i Y and Special Projects Office of Nuclear Reactor Regulation t

Enclosure:

As stated

a .

. . REQUEST FOR ADDITIONAL INFORMATION ADVANCED BOILIhG WATER REACTOR STANDARD 5AFETY ANALYSIS REPORT, DOCLET NO. 50-605 HECHANICAL ENGINEERIhG BRANCH 210.3 In Subsection 3.1.2.1 1.2, "Evaluation Against Criterion 1", a footnote states that "important-to-safety" and "safety-related" are considered equivalent in this SSAR. The staff does not agree with this definition. The staffs' position on this issue remains as stated in NRC Generic Letter 84-01, "NRC Use of the Terms  :

"important to Safety" and "Safety-Related", dated January 5, 1934.  !

,, The staff used this position as guidance in its reviews of applica- -

tions for operating licenses of nuclear power plants for a number l

j of years prior to the issuance of GL 84-01. During these reviews, the staffs' evaluations of the quality assurance requirements ir,

, 10 CFR Part 50. Appendix 8 generally applied to the narrower class l of"safety-related"equipmentasdefinedin10CFRPart50.49(b)(1),

i  :

] 10 CFR Part 100, Appendix A and in Section 3.2 of this SSAR. This l implied that nonnel industry practice for quality assurance was generally acceptable for most equigeent not covered by the "safety l

related" definition. However, as pointed out in Generic Letter 84-01, there have been specific situattoris in the past where the i staff has determined that quality assurance requirements beyond r

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normal industry practice were needed for components and equipment in i the more broso "important to safety" class.

It is the staffs' opinion that a strict interpretation of the ABWR position on this issue could result in an unacceptable classification i of structures, systems and components for Table 3.2-1 in this SSAR. [

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. . Revise the footnote in Subsection 3.1.2.1.1.2 and the discussion in Section 3.2 to be consistent with the staff's position as stated in Generic Letter 84-01. It should be made clear that the staff's position will not result in a broadening of the staff's review.

Rather, it provides the basis which the staff has been using and continues to use as guidance in its reviews of Quality Group Clas-sification for certain components and equipment which are not included in the "safety-related" definition.

210.4 In Subsection 3.2.3, "Safety Classifications". ANSI /ANS 52.1 - 1983, "Nuclear Safety Criteria for the Design of Stationary BWR Plants" is referenced for the definitions of safety classes. The guidance in this document for components which are not within the scope of Regulatory Guide 1.26 has not been endorsed by the staff. There-fore, the staff does not completely accept ANSI /ANS 52.1 for the definitions of all safety classes. Questions 210.5, 210.13, 210.15, 210.17, 210.44, and 210.45 are based on this position. To assure that Table 3.2-1 will be consistent with similar tables in recently licensed BWR/6 plants, such as Perry and River Bend, the reference to ANSI /ANS 52.1 - 1983 should be either eliminated or reviseo.

210.5 In Table 3.2-1. Items Bl.7, "Control Rods" and 81.9, "Fuel Assemblies" are classified as Safety Class 3, which is consistent with the cri-teria in the ANSI /Ah5 52.1 - 1983 Standard. As stated in Question 210.4 the staff does not agree with all of the recomendations in

-3 that Standard. The staff position is that Control Rods and Fuel Assemblies should be Safety Class 2 and Quality Group B. To be consistent with this position and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.2-1 to change the classifications of the Control Rods and Fuel Assemblies from Safety Class 3 to 2 and add Quality Group B.

Questions 210.44 and 210.45 provides similar staff positions for Item Bl.5 Safety-Related Reactor Internal Structures and Core Support Structures.

210.6 In Table 3.2-1, Jtem B2.5 identifies Main Steam Line (MSL) piping from the outermost isolation valve to and including the seismic interface restraint as being Safety Class 1 and Quality Group A.

Figure 5.1-3b, "Nuclear Boiler System P&lD Sheet 2" identifies the same portion of the MSL as Quality Group B. Beyond the seismic

interface restraint, the MSL piping is Quality Group D, which is not

! acceptable to the staff. To be acceptable, the MSL should be classifieo as recomended in Standard Review Plant 3.2.2, "System Quality Group Classification". Appendix A, i.e., Quality Group B from the outerinost isolation valve to the turbine stop valve. This staff position is based on the assumption that the ABWR MSL design differs from the BWR/6 design in that it does not contain a shutoff valve in addition to the two containment isolation valves. Revise Table 5.1-3b Table 3.2-1. Subsection 3.9.3.1.3 and Subsection 5.4.9.3 to be consistent with the above staff position.

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210.7 Item B2.5 in Table 3.2-1 does not appear to agree with Figure 5.1-3c, "huclear Boiler System P&!D. Sheet 3". Item B2.5 states that piping in the Feedwater (FW) Systems from the outermost isolation v61ve to and including the seismic interface restraint is Safety

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Class 1 and Quality Group A. Figure 5.1-3c shows the FW line es Quality Group A up to the first spring closing check valve outside containment (F262A). The FW piping is Quality Group B between valves F262A and F282A and Quality Group D beyond F282A. There does not appear to be a seismic restraint in Figure 5.1-3c. Assuming that the ABWR FW line is similar to the BWR/6 designs, i.e., valve F282A is a shutoff valve in addition to the two containment isolation valves, the Quality Group classification of this line does not appear to be consistent with the guidelines of Standard Review Plan  !

3.2.2, Appendix 8. Revise Table 3.2-1. Figure 5.1-3c and Subsection 5.4.9.3 to be consistent with the staff position on Quality Group in SRP 3.2.2, Appent x B. The transition from Quality Group B to 0 L should be at the seismic interface restraint rather than shutoff I valve F282A, I l

r 210.8 in Table 3.2-1, Item 83.1, the primary side recirculating motor cooling system piping is classified as Safety Class 3 and Quality Group C. In Subsection 3.9.3.1.4, this piping is described as being designeo to the ASME Code Section 111. Subsection NS-3600, which is comparable to Safety Class 1. In Figure 5.4-4, "Reaccor Recirculation System P&l0", this piping is identified as Quality Group A. The staff's position is that this piping should be, as a minimum Safety Class 1

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5-Quality Group A and meet the requireNnts of 10 CFR 50, Appendix B from the interface of the piping with the pump motor casing to and including the first pipe support. The remainder of this piping should be as a minimum Safety Class 2. In addition, item B3.2, the supports for this piping, shoulo be the same Safety Class as the supported piping. Revise items 83.1 and B3.2 in Table 3.2-1 to be consistent -

with the staff position.

210.9 in Table 3.21, add the classification summary for the Control Rod Drive Mechanism and the Low Pressure Core Flooder System or provide a justification for not including this information. The staff position on the Safety Class of these systems is as stated in  !

Questions 210.5 and 210.45.

210.10 Provide the basis for all Control Rod Drive System valves (Item C1.1 in Table 3.2-1) to be classified as Non-Nuclear Safety and Non-Seismic.

210.11 Provide the basis for portions of piping systems within the outer-most isolation valves in the Residual Heat Removal System and the Reactor Core Isolation Cooling System (Items E1.3, E4.1, and E4.6 in Table 3.2-1) to be classified as Safety Class 2 and 3.

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210.12 Items E2.1 and E2.5 in Table 3.2-1 classifies some pumps and valves l

within the outennost isolation valves in the High Pressure Core Flooder System as Safety Class 2. Provide the basis for this

! classification.

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210.13 In table 3.2-1, item F4.1, "Refueling Equipment Platform Assembly"

! is classified as Non-Nuclear Safety. To be consistent with the staff position as stated in Question 210.4 and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.2-1 to change this classification to Safety Class 2 and Quality Group B.

l 210.14 If a Fuel Transfer System or Tube is applicable to the ABWR, add the Classification Sunnary for this system under item F4, "Refueling Equipment" of Table 3.2-1, 210.15 In Table 3.2-1, Jtems FS.1, "Fuel Storage Racks-New and Spent" and F5.2, "Defective Fuel Storage Container" are classified as Non-Nuclear Safety, item FS.2 is also classified as Non-Seismic. To be consistent with the staff position as stated in Question 210.4 and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.2-1 to change the classification of Items F5.1 and F5.2 to Safety Class 3 and Quality Group C. In addition, change the seismic classification of item F5.2 to Seismic Category I and add "B" in the Qu611ty Assurance column for F5.2.

, 210.16 In Table 3.2-1, the following components in the Reactor Water C1vanup System are correctly classified as Quality Group C, but are

, also classified as Non-Nuclear Safety:

G1.1 - Vessels G1.2 - Regenerative Heat Exchanges 61.3 - Cleanup Recirculation Pump G1.5 - Pump suction and discharge piping beyond containment isolation valves.

G1.8 - Non-regenerative heat exchanger tube inside and piping and valves carrying process water.

G1.11 - Filter demineralizer holding pumps, valves and piping.

To be consistent with the discussiuns in Subsections 3.2.2 and 3.2.3 and with the information in T&bles 3.2-2 and 3.2-3, the staff is of the opinion tha*. all of the above components should be classified as Safety Class 3 in addition to Quality Group C. Revise Table 3.2-1. Items Gl.1, Gl.2, G1.3, G1.5, G1.8, and G1.11 to change the Safety Class from "N" to "3" or provide'a justification for not doing so.

210.17 In Table 3.2-1. Items G2.3, "Heat Exchangers", G2.4, "Pumps and Pump

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Motors" G2.5, "Piping, Valves", and G2.7 "RHR Connections" in the fuel Pool Cooling anc Cleanup System are all classified as Non-

e Nuclear Safety, which is consistent with the criteria in the ANSI /

ANS 52.1 - 1983 Standard. As stated in Question 210.4 the staff does n'ot agree with all of the recorsnendations in that Standard.

The staff position is that all of the above items should be Safety Class 3. Seismic Category 1 and listed under Quality Assurance requirements of 10 CFR 50, Appendix B. Regulatory Positions C.2 in R(gulatory Guide 1.26 and C.1 in Regulatory Guide 1.29 includes this 4

position. To be consistant with this position and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table l 3.2-1 to change the classification of items G2.3, G2.4 G2.5, and G2.7 from Non-Nuclear Safety to Safety Class 3, add Seismic Category 1 and ado "B" under Quality Assurance Requirement.

210.18 A staff position is that piping and valves forming part of primary containment boundary should be Seismic Category 1. In table 3.2-1, piping and valves in the Reactor Building Cooling Water System  ;

which fonn part of the primary contairant boundary are classified as Non Seismic. Revise Table 3.2-1 to add Seismic Category 1 to the l l

classification of item P2.1 or provide a justification for not doing l so. t 210.19 In Table 3.2-1, ths following items are classified as Seismic Category 1 without & connitment to the Quality Assurance Requiren.ent:

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. . B3.1 - Reactor Recirculation System piping, primary side, motor cooling.

F4.1 - Refueling equipment platform assembly.

F5.1 - Fuel storage racks, new and spent.

The staff position, as discussed in Posittor C.1 and C.4 of Regulatory ,

Guide 1.29 is thet quality assurance requirements of 10 CFR 50, Appendix B should be applied to all structures, systems and compo-nents which are classified as Seismic Category 1. Revise Table 3.2-1 [

to add "B" in the Quality Assurance Requirement column for Item B3.1, F4.1, and F5.1.

210.20 One of the staff positions relative to component supports is that the Safety Class Quality Group, Quality Assurance and Seismic Category classifications shall be identical for the supports and the  :

supported component. Provide a commitment to this position in  !

Table 3.2-1 and, if applicable, in Subsection 3.9.3.4, "Component Supports".

210.21 In Subsection 5.2.1.1. Table 3.2 4 is referenced to show the ABWR compliance with the rules of 10 CFR 50 Codes and Standards. Sub-section 3.2 in the SSAR does not contain a reference to Table 3.2-4 In either Subsection 3.2 or 5.2.1.1, provide the information re-quested in Standard Review Plan, Section 5.2.1.1, "Compliance With ,

the Codes and Standards Rule, 10 CFR 50.55a". This information i i

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, , should include the component Code Code Edition and Code Addenda which will be applicable to ABWR pressure vessels, piping, purps, valves, tanks, component supports and equipment.

210.22 Regulatory Guide 1.151 "Instrument Sensing Lines", dated July, 1983 conditionally endorses the Instrument Society of America Standard ISA-567.02, "Nuclear Safety-Related Instrument Sensing Line Piping and Tubing Standards foi Use in Nuclear Power Plants " 1980 as a basis acceptable to the NRC staff for the design and installation of safety-related instrument sensing lines in nuclear power plants.

In addition to the comitment in Table 1.8-20. provide a statement in either Section 3.2 or 3.9 of the SSAR, that the design of safety-related instrument lines for the ABWR will be in conformance with Regulatory Guide 1.151. Footnote g to Table 3.2-1 is related to this issue, but does not provide an explicit comitment to R.G.

1.151, 210.23 Subsection 3.6.1.1.3(2) states that a pipe break event will not occur simultaneously with a seismic event. This does not agree with Standaro Review Plan, Section 3.6.1, Branch Technical Position ASB 3-1, Paragraph B.2.b(1) or with the staffs

• interpretation of Plant Event 8 in Table 3.9-2 of the SSAR. Revise Section 3.6.1.1.3(2) to be consistent with the staff position in SRP 3.6.1 or provide a justification for not doing so.

. 210.24 The discussion in Subsection 3.6.2.2.1 (a) through (e) relative to the methodology used to determine blowdown forcing functions requires more detailed infortn. tion. Either revise this subsection to provide a consnita=r t to the non-n.endatory Appendix B of ANS 58.2 "Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Ruptures", or provide the fo119 wing:

a. Provide a detailed discussion of the basis for the 0.7thrustcoefficientinSubsection3.6.2.2.1(c).

b. In Subsection 3.6.2.2.1 (e) provide a discussion (iricluding references) of the methodology used tu reduce the thrust coefficient factors of 1.26 and 2.0 by accounting for friction.

210.25 Subsection 3.6.2.3.3 states that piping integrity does not depend on pipe whip restraints for any piping design loading combination including earthquake. Subsection 3.2.1 states that pipt whip re-straints need not remain functional in the event of a Safe Shutdown Earthquake. The staff agrees that pipe whip restraints do not have to be classified as Seismic Category 1. however, they should be designed to remain functional during a seismic event. Provide assurance that pipe whip restraints and their supporting structure cannot fati during a seismic event. If Subsection 3.8.3.3.2 is applicable to pipe whip restraints as well as their supporting

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l l structures, provide a reference to this Subsection in Subsection l 3.G.2.3.3. Revise Subsections 3.2.1 and 3.6.2.3.3 to be consistent with the response to this question.

210.26 In Subsections 3.7.2.1.3, 3.7.3.3.1.3, and 3.7.3.8.2.1, the rnultiple support excitation analysis method is referenced as an alternative to the envelope response spectrum method when calculating inertial l

responses of multiply-supported piping and equipment. This alternate method is acceptable to the staff only under the following conditions:

l l a. The multiple support input response spectrum method may be used only when support group responses are combined by the absolute sum method.

b. The multiple support input response spectrum method may not be used in analyses which also use the damping I values from ASME Code Case N-411. "Alternate Damping Valves for Seismic Analysis of Classes 1, 2, and 3 Piping Sections, Section !!!, Division 1." This position is one of the conditions listed in Regulatory Guide 1.84, Revision 24 for usir] Code Case N-411.

Provide a comitment to the above conditions in an appropriate Section in the $5AR and cross reference this comitment in Subsection 3.7.2.1.3, 3.7.3.3.1.3, 3.7.3.8.2.1 and any other subsection which discusses the multiple support ex1tation analysis alternative.

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, 210.27 The information in Subsection 3.7.3.4, "Basis of Selection of Frequencies" does not appear to be consistent with the guide-

. lines in Standard Review Plan, Section 3.9.2 Pe.ragraph !!.2.C.

Revise Subsection 3.7.3.4 to include a comitment that, to  !

avoid resonance, the fundamental frequencies of components and l l

equipment should be selected to be less than 1/2 or more than twice the dominant frequencies of the support Structure.

. " 210.28 In Subsection 3.7.3.10, the statement is made that the vertical <

ground design response spectrum is used for equipment vertical seismic load deterinination if it can be showr that the structures supporting the equipment are rigid or quasi-?igid in the vertical direction. Provide definitions of "rigid, 'quasirigid" and "sup- l l port structure" in Sub-section 3.7.3.10, 210.29 Subsection 3.9.2.2.2.1 states that prelimirary dynamic tests are '

conducted to verify the operability of the control rod drive (CRD)  !

during a dynamic event. Provide a more detailed description of i these tests and, if applicable, discuss htw the results of the

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i testsarecorrelatedwiththeabalysisoftheCRDhousing(with h the enclosed CRD) which is mentioned in tne first sentence of this subsection. If the fine motion control rod drive system is not

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included in these tests, describe how that system is seismically  !

qualified. '

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. 210.30 Revise the discussion Subsection 3.9.1.4.4 to be consistent with the information in Subsection 3.9.3.4.3 for the reactor pressure, vessel stabilizer and Subsection 3.9.3.5 for the supports for the fine enotion control rod drive and in-core housings.

210.31 In Subsection 3.9.2.1.1. ANSI /ASME OM3 1987, "Requirements for Preoperational and Initial Startup Vibration Testing of huclear Power Plant Piping Systems" is referenced for vibration testing of ABWR piping systems. However, in Subsections 3.9.2.1.2 and 14.2.12.1. there is no reference to OM3 for preoperational themal expansicn and dynamic testing and the information in these sub-sections on these phases of preoperational testing is not presented in sufficient detail for the staff to evaluate. Revise Subsections 3.9.2.1.2 and 14.2.12.1 to aither include a reference to ANS1/ASME OM3-1987 or present inforntion similar to that for the Main Steam Line piping which is discussed in Subsections 3.9.2.1.3, 3.9.2.1.4 3.9.2.1.5 and 3.9.2.1.6.

210.32 In Subsections 3.9.2.1.1 and 14.2.12.1. there is no mention of preoperational vibraticn testing of safety-related instruuntation lines. It is the staff's position that all essential safety-related instrur.entation lines and sm:.11 bore piping should be included in the vibratior, monitoring program during preoperational or start-up testing. We require that either a vis al or instrumented inspection (as appropriatt) be conducted to identify any excessive vibration that could result in fatigue failure. Generally. this includes the

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portion up to and including the first support away from the connection to large bore piping or component. If observations suggest that other spans are being excited, further inspection would be conducted on a case by case basis. Revise the above Subsections to provide a commit-ment to this position.

210.33 The discussions in Subsection 3.9.2.5 and 3.9.5.2 relative to the dynamic system analysis of reactor internals under faulted conditions does not provide enough detailed information for the staff to evaluate.

Standard Review Plan, Section 3.9.2.!!.5 provides the acceptance criteria which the staff uses to evaluate this issue. Information in sufficient i detail to implement this criteria is required before the staff can complete its evaluation. Revise Subsection 3.9.2.5 to include this information either in the form of references or an additional appendix in Section 3-2 of the ABWR SSAR.

210.34 In Table 3.9-2, the acceptance criteria for the stresses resulting from the service loading combination of rorwul loads plus the r.ost I'miting safety / relief valve loads plus turbine stop valve closure induced loads is identified as ASME Level D Service Limits. If this is a typographical error, replace Level D with Level 6 in this table. If it is not en error, provide the justification for using Level D Service Limits for this loading combination.

e 210.35 Provide the basis fy assuring that the feedwater isolation check valves can perform its intended function and satisfy GDC 54 and 55 following a feedwater line break outside ,

containment. Aoditionally, discuss what actions have been taken to preclude the possibility of a feedwater pump trip transient causing a feedwater line break outside containment.

210.36 The discussions of ASME Class 1, 2 and 3 safety-related code components in Subsections 3.9.3.1.3 through 3.9.3.1.7 and 3.9.3.1.9 through 3.9.3.1.19 use the terms "designed and evaluated" in accordance with ASME Section !!! rules for Class j 1, 2 and 3 components. In discussions of this nature, the word "constructed" should be used rather than "designed and evaluated" where construction is defined in accordance with the ASME Section

!!!. Subsection NCA 1100 definition, i.e., "an all inclusive tem comprising materials, design, fabrication, examination, testing, inspection and certification required in the manufacture and installation of items". Revise all of the above Subsections to state that all of these components are constructed in accordance with the ASME !!! NCA 1100 definition.

210.37 Subsection 3.9.3.2 contains several references to IEEE-344. "!EEE Recorvr. ended Practices for Seismic Qualification of Class IE Equip-ment for Nuclear Power Generating Stations" with no issue date.

, To be consistent with current staff positions on this issue, revise each of these references to read "IEEE STO. 344-1987" and add a commitment to NRC Regulatory Guiot 1.100, Revision 2 "Seismic Quelification of Electrical Equipment in Nuclear Power Plants" to each reference. The staff considers these two documents to be applicable to mechanical as well as electrical equipment.

210.38 Subsectior. 3.9.3.3.2, "Other Safety / Relief Valves" references ASME Section !!!, Appendix 0 for the safety-relief valve opening and pipe reaction loads which will be used in the design of ABhR safety-relief valves. The staff's position on this issue is that if Appendix 0 is used, the additional criteria ir, Standard Review Plan. Section 3.9.3, Paragraph II, 2 is applicable. Revise Subsection 3.9.3.3.2 to include a commit-ment to this position.

210.39 Subsections 3.9.3.4.1 and 3.9.3.5 both state that the jurisdictions 1 beundary between component supports designed to ASME Section !!!, Subsection hF and the building struc-tive shall be as defined in the project design specifications.

The project design specifications may or r.ay not agree with the definitions of jurisdictional boundaries which are in ASME Subsection NF. Therefure, revise Subsections 3.9.3.4.1

. . and 3.9.3.5 of the ABWR SSAR to provide a commitment that the 1987 Adder.da to the 1986 fdition of ASME Section !!!. Subsection NF will be used to define the jurisdictional boundary between Subsection NF component supports and the builoing structure.

210.40 The information in Subsections 3.9.3.4.2 and 3.9.3.5 relative to analyses for buckling of the reactor pressure vessel support skirt and other ASME !!! component supports needs to be updated ano clarified as follows:

a. Paragraph 1370 (c) of ASME !!! Appendix F which is referenced in both of the above subsections was deleted in the Summer,1983 Adoenda to ASME Ill Division 1 Appendices. ASME Appendix XVII, which is referenced in Subsection 3.9.3.5 was deleted in the Winter. 1985 Addenda. Revise Subsections 3.9.3.4.2 and 3.9.3.5 to provide references which are applicable to the latest l edition of ASME Section !!!.

l l b. Provide a more detailed description of how the critic 61 l

buckling strength of the RPV support skirt and other A5ftE !!! component supports will be determined.

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t 210.41 The following additional information is required in Subsection 3.9.3.4 e

relative to the design of bolts for component supports: t

1. Provide the allowable stress limits and/or safety factors

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which are applicable to bolts used in equipment anchorage.  !

component supports and flanged connections.

Specifically provide a discussion of the design methods applicable to expansion anchor bolts and case-in-place used in component supports and equipment anchorage.

210.42 In Subsection 3.9.3. provide the design basis which will be used in the ABWR to insure the structural integrity of safety related heating, ventilation and air conditioning ductwork and its supports.

210.43 Subsection 3.9.4 outlines seven types of tests which will be used as a basis for the ABWR Control Rod Drive (CRD) Perfor-mance Assurance Program. The first type. ' Development Tests" are discussed in Subsection 4.6.3.1. According to this dis-l cussion, at least three different prototype designs of the FineMotionControlRodDrive(FMCRO)havebeensubjectedto various test programs. The staff's Question 440.8 requested l

l the results of the tests of the inplant FMCR0 prototype which

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are currently being conducted at La Salle, Unit 2. In addition to a response to Question 440.8, provide a description of the differences between the initial, inplant and reference FMCRD designs and, if applicable. 4 discussion of any correlation that may exist between the accumulated test data from all three designs and the design criteria discussed in Subsections 3.9.1.1, 3.9.1.4 and 3.9.3 and Table 3.9-2, 1

210.44 Subsection 3.9.5.1.1 states that the core support structures in the ABWR are classified as Safety Class 3. The steff's position is that (

these structures are necessary to help maintain core geometry and should therefore ce classified as Safety Class 2 to obtain a higher level of quality assurance than Safety Class 3. Revise Tables 3.2-1 h and 3.2.3 and Subsection 3.9.5.1.1 to agree with this position.  :

1 210.45 In Subsections 3.9.5.1.2.4, 3.9.5.1.2.5 and 3.9.5.1.2.6 the t

feedwater spargers, RHR/ECCS low pressure flooder spargers and the ECCS high pressure core flooder spargers and piping are all classifieo as Safety Class 3. The staff's position is that these [

r reactor internal components are necessary to help accomplish the safety function of veergency core cooling and should therefore be classified as Safety Class 2 to obtain a higher level of quality f

essurance than Safety Class 3. Revise Table 3.2-1 and Subsections L 3.9.5.1.2.4. 3.9.5.1.2.5 and 3.9.5.1.2.6 to agree with this position.

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1 210.46 Portions of the stress, deformation and buckling limits for safety class reactor internals which are listed in Tables l 3.9-4, 3.9-5 and 3.9-6 requires additional review by the i

staff. If either Equation b in Table 3.9-4 Equations e, f, and g in Table 3.9 5 or Equation c in Table 3.9 6 will be used in the design of safety class reactor internals for the ABWR, provice a comitment in each of these tables that supporting data will be provided to the staff for review.

J 210.47 The information in Subsection 3.9.6 infers that only ASME Class 1, 2 and 3 pumps and valves will be included in the inservice testing (IST) program for the ABWR. It is the i

staff's position as stated in Standard Review Plan, i

Sections 3.9.6.!!.1 arid 3.9.6.!!.2 that all pumps and i'

valves which are considered 45 safety-related should be included in the !$T program even if they are not cate-gorized as ASME Class 1, 2 or 3. Revise Subsection 3.9.6 to agree with this position.

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210.48 The first paragraph in Subsection 3.9.6 states that accessi-bility for inservice testing of applicable purps and valves is provided in the plant design. However, the second paragraph and Subsection 3.9.6.3 infers thet relief from ASME Section XI inservice testing will be submitted for some pumps and valves.

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All of the plants which have been licensed by NRC so far have been allowed to request relief from the ASME Section XI inservice testing rules for a limited number of pumps and velves. 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 purrp or valve difficult to test without additional design changes. Therefore, the staff granted many of these requests for relief because imposition of these rules would have resulted in hardships to the licensve without a compensating increase in i 1

the level of safety. The underlying reason for the regulation

, allowing tnese reliefs from the code was that the detailed l piping system designs for all of these plants was completed i prior to the tire that the staff began to implement the ASME Section XI rules.

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A plant such as the ABWR. for which the final design is not complete, has sufficient lead time available to include provisions 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 arid valves will not be granted for the ABWR. Revise Subsection 3.9.0 to provide a more explicit comit.

meht that ABWR piping systvais will be designed to accomodate the applicable code requirements for inservice testing of pumps ano valycs. However. with regard to subsequent or future code revisions

e6 O 23 to the applicable ASME Code for the ABWR plants, requests for relief from certain updated code requirements may still be submitted for staff review in accordance with 10 CFR 50.55a(g).

210.49 In Subsection 3.9.6, "Inserivce Testing of Pucps and Valves "

provide a comitJnent to perform periodic leak testing of all pressure isolation valves in accordance with the applicable sections of the Technical Specifications for recently licensed BWR/6 plants. Normally, this information includes a list of all pressure isolation valves which will be leak tested, if such a itst is not available for the ABWR, a consnitment to provide the list of valves as a part of the ABWR Technical Specifications will be acceptable.

210.50 In accordance with hRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems " the staff is currently requesting licensees and applicants to review systems connected to the reactor coolant system to detennine whether any sections of such j piping which cannot be isolated can be subjected to stresses from ,

temperature stratification or temperature oscillations that could be induced by leaking valves. If this phenomenon was not considered in {

the design analysis of the ABWR piping, submit a response to action [

ltem 3 in Bulletin 88-08 which will be applicable.

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Seismic and dynamic load qualification 271.01 Subsection 3.10.1.3 states that the ABWh program for dynamic qualification of Seismic Category 1 electrical equipment meets the criteria contained in IEEE-344 as modified and endorsed by Regulatory Guide 1.100. To be consistent with recent staff positions on this issue, revise Subsection 3.10.1.3 f 1 -

• E-344-1987 as modified and endorsed by Regulatoi de 1.100, Revision 2".

271.02 Subsection 3.10.1.3, "Dynamic Qualification Program" states that Section 4.4 of GE's Environmental Qualification Program (NEDE-24326-1-P) will be used for dynamic qualification of Seismic Category 1 electrical equipment and that this report is referenced in Subsection 3.11. The reference in Subsection 3.11.7 is to the January, 1983 version of NEDF,24326-1-P.

The staff's approval of this report is based on the January, 1986 Kavision. Revise Reference 2 in Subsection 3.11.7 to change the date of NEDE-24326 from January,1983 to January,1986.