ML20004E174

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Forwards Responses to NRC Questions.Responses Will Be Filed in FSAR Amend
ML20004E174
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 06/09/1981
From: Colbert W
DETROIT EDISON CO.
To: Kintner L
Office of Nuclear Reactor Regulation
References
EF2-53-498, TAC-66314, NUDOCS 8106110300
Download: ML20004E174 (150)


Text

{{#Wiki_filter:~ 86g3['Eg 4 Detroit Edison E75=6 1 ,8-June 9, 1981 th 8g EF2 - 53,498 \\C \\ Mr. L. L. Kintner Division of Project Managenent Office of Nuclear Regulation U. S. Nuclear Regulatory Ccmnission Washington / '). C. 20555

Dear Mr. Kintner:

Reference:

Enrico Fermi Atcmic Power Plant Unit 2 NRC Docket No.'50-341

Subject:

Detroit Edison Responses to NRC Ouestions Please find enclosed several Detroit Edisen responses to open NRC questions. These responses will be filed in a forthccming FSAR amendment as appropriate. Item 1 CPB Verbal - B.K. Sun CPR Limits Detroit Edison's response to Dr. Sun's verbal question on minimum cperating CPR limits is enclosed as Attachment 1. Item 2 RSB Savannah River LPCI Divers 13 Detroit Edison's response to this item is enclosed as Attachment 2. Item 3 CES C021.33 Fire Protection Detroit Edison's response to this question is enclosed as Attachment 3. g e Dk 'p THIS DOCUMENT CONTAINS 8106 110 @ P00R QUAUTY PAGES 6

June 9, 1981 Mr. L. L. Kintner EF2 - 53,498 Item 4 Appendix J, Tyoe C Tests CSB Detroit Edison has previously amended Table 6.2-2 in Amendments 29 and 33. Please refer to thesa amendments to clarify (.mt confusion over status of this table. Enclosed as Attachment 4 is a table identifying exceptions to 10 CFR 50 Appendix J as requested by NRC staff. Item 5 MEB June 4, 1981 Questions Supolenental Seisaiic Peport Detroit Edison's response to the four MEB questions on the Supplemntal Seismic Report are enclosed as Attachment Sa and Sb. Item 6 ICSB Ve rbal Questions - Mauch Various Please find eaclosed as Attachmnt 6 our responses to several verbal questions dated June 9, 1981, frcm Gerry Mauch. These responses address: e FSAR 7.3.9, RIND I&C e FSAR 7.4.2.4, Reactor Shutdown Cooling I&C e RCIC Auto Suction Switchover Irgic Item 7 Revised FSAR Pages As a result of responses to other N7f questions, Detroit Edison has revised Pages 9A-5, 9A-6, 10.3-1,

  • .,.! 10.3-2 of the FSAR. These revisions are enclosed as Attachmr at 7.

Item 8 ASB 0020.26 MSIV Testing Internal and Iaakage Rates Detroit Edison's response to this question is enclosed as Attachment 8. --m-g - ~~ n,,, ,~,,--,--,,.,, -,, w --~u -n

June 9, 1981 Mr. L. L. Kintner EF2 - 53,498 Item 9 RSB Savannah River Mwiual Actions - IOCA Detroit Edison's response to this item is enclosed as Attachment 9. Sincerely, William F. Colbert Technical Director Enrico Femi Unit 2 WFC:RMB/cm Attachments: As Ibted Above 4 i J u....

June 9, 1981 Mr. L. L. Kintner EF2 - 53,498 bec: R. M. Berg F. E. Gregor J. W. Honkala E. Lusis L. E. Schuer:ran A. E. Wegele Document Control J

d'TT)4CMM d SFa s3nt ADDITIONAL INFORMATION ON MINIMUM OPERATION CPR LIMITS ,: e This additional information is being provided in response to informal questions raised by the NRC staff with respect to operating CPR limits and Tech. Spec. commitments.

RESPONSE

a. Figure 15B-0-6 of the FSAR will be revised to indicate LFWH line as dashed line instead of a solid line as currently shown in FSAR. (See attached Figure 15B-0-6 revised) b. Figure 2 is being provided to indicate new operating limit CPR (TT (LR) with no BP with RH). This curve represents the new ODYN analysis for Turbine Trip (Load Rejection) case with no bypass flow but with reheater flow. c. Section 3/4-2.3 " MINIMUM CRITICAL POWER LIMITING CONDITION FOR OPERATION" of the Tech. Spec. will include the MCPa as a function of scram time as shown in attached Figure 2. N. Kt Deora /dk -s 6-9-81 f G B I

EF-2-FSAR G 8 N7 l i i n I 1 6 \\ s = ,\\

== \\ I = ;: b< I I \\ ~ oz \\ \\ 25 I as Ew I \\ I \\ i l I \\ \\ I \\\\ 3 i j \\ \\. i \\ \\ _~ I 1 \\ b I-s I i m l 2 1 = i g 6 [I EE 55 = E b! 55 EE D I 3 3, cm I g y p_= 5 I 1 t e ~ 8 ~8 4 8 E 9 E IlWl18d3 DN11VH3d0 W!1WINIPi ~ ~. ENRICO FERMI ATOMIC POWER PLANT UNIT 2 ENVIRONMENTAL REPORT FIGURE 158.0-6 MINIMUM OPER ATING CPR Li'itT VERSUS SCR AM SPEED AMEt4DMENT 24 - APRIL 1931 153.0-27

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ATTACMM A EfD-C'fW S. R. QUESTION 6, RESPONSE LOCA nNALYSIS (LPCI DIVERSION) The SER states that "if controls, stronger than procedural controls, are not used to prevent LPCI diversion prior to 10 minutes, then analyses supporting earlier diversion are required." Analyses of BWR performance following a small break LOCA and LOCA mitigatiot. under degraded conditions have been performed by General Electric as a part of the BWR Owners Group program. Analyses bases, assumptions and conclusions are discussed in General Electric Report NEDO-24708A, Revision 1, December 1980 entitled, " Additional Information Required For NRC Staff Generic Report on Boiling Water Reactors." Reference is made to Section 3.1.1 (small break LOCA) and 3.5.2 (inadequate core cooling). It should be noted that these analyses were performed utiliz-ing " realistic" assumptions as defined in Sections 3.1.1.3 and 3.5.2.4. The conclusion Euction 3.5.2.1.8 summarizes the capability of the BWR to maintain adequate core cooling, even under severely degraded ccnditions resulting from multiple failures and operator errors, following a loss of inventory either through a pipe break or through the safety relief valve. Based on the first group of analyses presented it was con-cluosd that for any plant and any loss of inventory event, the availability of ADS and one low pressure ECC system pro-vides adequate core cooling if no high pressure injection is =

LOCA ANALYSIS (LPCI DIVERSION) PAGE 2 available. These analyses covered the case of multiple mechanical or electrical failures and operator errors that might have caused the failure of the system, assumed to be unavailable. The second set of analyses addressed the condition of the vessel being at high pressure with a low wa*er level. It was shown that operator actions either to initiate high pressure systems or to depressurize the vessel and initiate at least one low pressure system, terminate this condition and assure adequate core cooling. The analyses showed that even for such severely degraded transients there is sufficient time for operator action to mitigate the consequences. The third set of analyses address the condition of the vessel being at low pressure with a low water level but with the low pressure systems not injecting. It was shown that opera-tor actions either to start the low pressure systems injecting into the vessel or to initiate the high pressure systems, termi-nate this condition and assure adequate cote cooling. l For all analyses, it was shown that the process variable in-formation available to the operator in the control room is sufficient to adequately warn cf an inventory threatening evert and to present the information the operator needs to assure that appropriate actions are taken to maintain adequate core coolina. The control room indications will not mislead the operator when taking corrective actions. Even under the 4 extremely degraded conditicns considered in these analyses, t

LOCA ANALYSIS (LPCI DIVERSION) PAGE 3 the BWR requires only the most basic opeator actions to mitigate the consequences of an inventory threatening event. If the operator were to divert LPCI prior to 10 minutes post-LOCA, such an action would be considered an operator error. Since the current ECCS performance evaluation already assumes the accident, a loss of offsite power and a worst active single fil. lure, an additional operator error is considered to be unnecessary additional Appendix K assumption. It is' therefore appropriate that the " realistic" assumption analysis be considered for this situation as stated in the conclusions in NEDO-24708A "for any plant and any loss of inventory event, the availability of ADS and one low pressure ECC system pro-vides adequate core cooling. This analysis is deemed acceptable to provide satisfactory assurance of acceptable event consequences, in consideracion of the equipment failures and operator errors assumed. Controls stronger than procedural are definitely not recom-mended to prevent LPCI diversion. F. K. Deora /dk 6-9-81 s

.A..: w - _w-. .:c _ w __ An* A T77H wiessC 3. L 6+a - syys/ /., * '. Et:Ct.05URE .~. >' .= 021.0 Chemical e naineerinc Branch - Fire Protection Review.' 2 021.33 Appendix R to 10 C.:R Part 50 will also be used as g id of your fire pectection program. ance fd our r'evisw u k license ccnditica.fcrth in Appendix P. as modified by accepted excep ~ ments of Appendix R as well as STP ASS 9.5-1 Identify any exc ns will be made a y 'r for providing an equivalent level of fire pro,tectiand describe your altarna re-h y on. I k RESPONSE 021.33 i A fire hazards analysis was conducted for the Enrico Fermi Unit 2 plant and a point-by-point comparison made with I Appendix A to BTP APCSB 9.5-1. This comparison is included in Appendix 98.5 of the FSAR. { The guidance of Appendix a to 10 CFR 50, as enumerated in the Federal Register / Volume 45, No. l 225, of November 19, 1980, has been reviewed. Enrico Fermi Unit 2 is in compliance ,I 1 t with the guidance of Appendix R as presented to the NRC staff in the May 27, 1981 meeting in Bethesda. l l 1, 1 tl l \\ R. C. Anderson a l /dk 6-4-81 lI!! a f f p 1 I i I -A ma u

TABLE '- EXCEPTIONS TO 10CFR50 APPENDIX J Penetration Valve Humber System Title Number Exception Justification X-7A Main Steam Line A V17-2003 Note 1 Note 2 X-78 Main Steam Line B V17-2001 Note 1 Note 2 X-7C Main Steam Line C V17-2002 Note 1 Note 2 X-7D Main Steam Line D V17-2004 Note i Note 2 X-8 Main Steam Line Drains V17-2009 Note 3 Note 4 X-10 Steam to RCIC Turbine V17-2030 Note 3 Note 4 X-Il Steam to HPCI Turbine V17-2020 Note 3 Note 4 X-13A RHR Pump Discharge to V8-2164 Note 9 Note 10 Recire Loop ,i

  • X-138 RHR Pump Discharge to V8-2163 Note 9 Note 10 Recirc Loop X-15 Combustible Cas Control V4-2144 Note 5 Note 4 System Suction X-16A Core Spray Pump Discharge V8-2024 Note 9 Note 10 X-16B Core Spray Pump Discharge V8-2023 Note 9 Note 10 X-17 RHR Discharge to Head Spray V8-2172 Note 3 Note 4 f

X-22 Nitrogen to Drywell V4-2080 Note 1 Note 2 kh X-23 Reactor Building Closed V8-2485 Note 3 Note 4 Cooling Water Supply \\}," cs S

TABLE - EXCEPTIONS TO 10CFR50' APPENDIX J Penetration Valve Number System Title Number Exception Justification X-24 Reactor Building Closed V8-3890 & Note 3 Note 4 Cooling Water V8-2486 X-25 Drywell Exhaust and Air VR3-3024 Note 5 Note 4 Purge X-26 Drywell Air Purge Inlet VR3-30ll Note 5 Note 4 L-27a Containment Atmosphere VS-2159 Note 6 Note 4 Sample X-27b Containment Atmosphere V5-2160 Note 6 Note 4 Sample X-27c Containment Atmosphere V5-2161 Note 6 Note 4 Sample e X-27d Containment Atmosphere V5-2162 Note 6 Note 4 Sample X-27e Containment Atmosphere V5-2163 Note 6 Note 4 Sample X-27f Containment Atmosphere V5-2164 Note 6 Note 4 Sample X-29Aa Reactor Water Sample V17-2077 Note 1 Note 2 X-29Be Drywell Instrumentation V5-2231 Note 6 Note 4 X-31B Drywell On-Line Pressure VR3-2825 Note 1 Note 2 Control X-34A Reactor Building Closed V8-2484 Note 3 Note 4 Cooling Water Supply i

i TABl.E - EXCEPTIONS TO 10CFM50 APPENDIX J Penetration Valve 3 Number System Title Number Exception Justification V X-34B Reactor Building Closed V8-3889 & Note 3 Note 4 Cooling Water Return V8-2483 X-36 Nitrogen to Drywell V4-2188 Note 1 Note 2 X-39A RilR to Containment Spray V8-2169 Note 11 Note 12 l i llende r X-398 RilR to Containment Spray V8-2170 Note 11 Note 12 lleade r X-44 Combustible Gas Control V4-2143 Note 5 Note 4 System Suction X-47e Drywell Pressure V5-2230 Note 6 Note 4 X-48a Containment Atmosphere V5-2151 Note 6 Note 4 Sample p X-48b Containment Atmosphere V5-2152 Note 6 Note 4 1 Sample F X-48c Containment Atmosphere V5-2153 Note 6 Note 4 Sample J X-43d Containment Atmosphere V5-2154 Note 6 Note 4 Sample X-48e Containment Atmosphere VS-2155 Note 6 Note 4 1 Sample t i lt' 1 X-48f Containment Atmosphere VS-2156 Note 6 Note 4 ( Sample (! 'l X-49a Recirc Pump Seal Purge V8-3767 & Note 3 Note 4 V8-3710 W i;! 5 j;i q, 1-

TABLE - EXCEPTIONS TO 10CFR50 APPENDIX J Pene t ration Valve Number System Title Number Exception Justification X-Sla Recirc Pump Seal Purge V8-3768 & Note 3 Note 4 V8-3590 X-204A Drywell to Torus Vacuum V4-2036 Note 1 Note 2 ] Breaker Nitrogen Supply X-204B Drywell to Torus Vacuum V4-2065 Note 1 Note 2 Breaker Nitrogen Supply { X-204C Drywell to Torus Vacuum V4-2075 Note 1 Note 2 4 Breaker Nitrogen Supply X-204D Drywell to Torus Vacuum V4-2077 Note 1 Note 2 Breaker Nitrogen Supply X-204E Drywell to Torus Vacuum V4 -2082 Note 1 Note 2 j Breaker Nitrogen Supply X-204F Drywell to Torus Vacuum V4-2084 Note 1 Note 2 Breaker Nitrogen Supply I X-204G Drywell to Torus Vacuum V4-2086 Note 1 Note 2 j Breaker Nitrogen Supply ? I X-204H Drywell to Torus Vacuum V4-2088 Note 1 Note 2 Breaker Nitrogen Supply 1 X-204J Drywell to Torus Vacuum V4-2090 Note 1 Note 2 Breaker Nitrogen Supply X-204K Drywell to Torus Vacuum V4-2092 Note 1 Note 2 Breaker Nitrogen Supply l j-X-204L Drywell to Torus Vacuum V4-2094 Note 1 Note 2 Breaker Nitrogen Supply s t I l j l

TABLE - EXCEPTIONS TO IOCFR50 APPENDIX J Penatratien Valve. Number System Title Number Exception Justification X-204M Drywell to Torus Vacuum V4-2096 Note 1 Note 2 Breaker Nitrogen Supply X-205A Torus to Secondary Contain-V21-2015 Note 5 Note 4 } ment Vacuum Breaker X-205B Torus to Secondary Contain-V21-2016 Note 5 Note 4 ment Vacuum Breaker X-205C Torus to Secondary Contain-VR3-3013 Note 5 Note 4 ment Vacuum Breaker X-20$D Torus to Secondary Contain-VR3-3015 Note 5 Note 4 ment Vacuum Breaker i X-206A Torus Pressure and Liquid V5-2552 Note 3 Note 4 Level Instrumentation X-206B Torus Pressure and Liquid V5-2553

1. Note 13 1

Vote 14 Level Instrumentation

2. Note 3 7 Note 4 X-206C Torus Pressure and Liquid V5-2550 Note 3 Note 4 Level Instrumentation h

X-206D Torus Pressure and Liquid V5-2551

1. Note 13
1. Note 14 i

Level Instrumentation

2. Note 3
2. Note 4 1

X-206E Torus Pressure and Liquid V5-2555

1. Note 13
1. Note 14 Levul Instrumentation
2. Note 3
2. Note 4 X-206F Torus Pressure and Liquid V5-2556
1. Note 13
1. Note 14 i

Level Instrumentation

2. Note 3
2. Note 4

.I i l i i 1 l 8

TABLE - EXCEPTIONS TO 10CFR50 APPENDIX J t.U Penatration Valve g. Number System Title Number' Exception Justification s )g-X-210A&B RilR All penetra-

1. Note 13
1. Nota 14 tion isola-
2. All valves
2. Note 4 or j!

tion valves tested in the results are jj. reverse conservative h direction since test pres-j.? sure tends to d I unseat the valve h disk X-211A RilR to Suppression Pool V8-2158 Note 1 Note 2 1 Spray 4 j[i X-211B RllR to Suppression Pool V8-2157 Note 1 Note 2 Spray g V. X-212, 214,& RCIC Turbine Exhaust Line Vll-2002 Note 1 Note 2 M 220 HPCI Turbine Exhaust Line Vll-2006 Note 1 Note 2 hh RCIC Vacuum Breaker Line Vll-2026 Note 3 Note 4 L L llPCI Vacuum Breaker Line Vll-2019 Note 3 Note 4 X-213A&B Torus Water Management All penetra-

1. Note 13
1. Note 14 Suction tion isola-
2. All valves
2. Note 4 or tion valves tested in the results are

- p reverse conservative h direction since test pres-I sure tends to unseat the valve l, disk h i X-215 Combustible cas Control V4-2142 Note 5 Note 4 h N System p X-215 Combustible Gas Control V5-2158 Note 6 Note 4 System p' '. p s

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TABLE - EXCEPTIONS TO 10CFR50 APPENDIX J Penetration Valvm Number System Title Number Exception Justification X-218 Combustible Gas Control V4-2140 Note 5 Note 4 System i X-218 Combustible Gas control V4-2139 Note 5 Note 4 System X-219 Combustible Gas control V4-2141 Note 5 Note 4 System I X-219 Combustible Gas Control V4-2166 Note 6 Note 4 System X-221 IIPCI Turbine Exhaust Drain Vll-2008 Note 1 Note 2 X-222 RCIC Vacuum Pump Discharge V8-2235 Note 1 Note 2 X-223A RilR Pump Suction All penetra-

1. Note 13
1. Note 14 through D tion isola-
2. All valves
2. Note 4 or the l

tion valves tested in results are d reverse conservative j direction since test pres-sure tends to un-3 seat the relief valve l X-224A&B Core Spray Pump Suction All penetra-

1. Note 13
1. Note 14 tion isola-
2. Note 3
2. Fate 4 tion valves 4

X-225 IIPCI Pump Suction All penetra-

1. Note 13
1. Note 14 4

tion isola-

2. Note 3
2. Note 4 q

tion valves X-226 RCIC Pump Suction All penetra-

1. Note 13

.l. Note 14 i tion isola-

2. Note 3
2. Note 4 tion valves l

I i n r A

TABLE - EXCEPTIONS TO 10CFR50 APPENDIX J Penetration Valva Number System Title Number Exception Justification I X-227A&B Core Spray, Torus Water All penetra-

1. Note 13
1. Note 14 Hanagement, HPCI, & RCIC tion isola-
2. All valves
2. Note 4 or the i

tion valves tested in results are reverse conse rva tive direction since test pres-sure tends to un-seat the relief ) valve X-230 PCMS and Postaccident V5-2157 Note 6 Note i Atmosphere Sample Suctione X-231 PCMS and Postaccident V5-2165 Note 6 Note 4 Atmosphere Sample Suctions j f i 1 4 4 4 } i l i l 1 l 1

Nates: 1. Globe valve tested in the reverse direction. 2. The results obtained in this test configuration are conservative since test pressure tends to unseat the valve disk. 3. Wedge-disk gate valve tested in the reverse direction. 4. The results obtained in this test configuration are equivalent to testing in the accident direction, since valves of this type have the same sealing characteristics in either direction. 5. Hutterfly valve tested in the reverse direction. 6. Ball valve tested in the reverse direction. 9. This valve will be seat leak tested with high pressure water in ac-a cordance with the requirements of ASME Section X1, Category A. 10. This valve fuctions as a high pressure (reactor recirculation system) isolation valve. Subject valves ata not required for the low pres-sure containment isola t ion func t ion. 11. Due to system configuration the test pressure is not in the same direction as the pressure existin-9 hen the valve is required to perform its containment isolation.unctio... 12. The valve will be tested in the correct direction during the Type A tests. i 13. This valve will be Type C seat leak tested using water as the test medium. [3 l m I m f E I

k l 14. The flow path associated with this penetration inside containment i terminates below the low water level in the suppression pool. A water seat is assured during normal plan

  • operation and for more 4

than 30 days following an accident requiring containment isolation. l It is not credible that these isolation valves will be exposed to the containment atmosphere at any time following the accident. l k l ,) i l; i F d a i s e i 1 1 b t .i i ) W l P l /' r. i

4Tucmev7-- n EFG -s'sy f) MEB QUESTIONS & RESPON3ES Question la: Provide the original design basis response spectra used in the main steam and reactor recirculation piping analyses. Answer la: The input OBE response spectra for the MS and RRS piping systems (' % damping) are attached (sheets s l A-1 through A-4). To obtain the equivalent SSE in-i put response spectra, multiply the OBE spectra (h%) by 2.0. l question Ib: Provide the site specific response spectra used for the MS and RRS piping in the siesmic reassessment

report, l

Answer Ib: The horizontal site specific response spectra in the reassessment report are as follows: j Steam Piping - Envelope of attached figures B53 & l B55. RRS Piping - Envelope of attached figures B47 & B48. l l The vertical site specific response was obtained by ( multiplying the existing SSE response by 2.0. The equivalent vertical response spectra is 4.0 times the input OBE response spectra (h%, sheet A-2 or A-3 attached). Question ic: Provide the site specific spectra applicable to the l MS and RRS support and anchor locations if a =ulti-j pie support analysis were performed. Answer ic: The following site specific response spectra are applicable to the RRS (and attached RRRS, RHRR) support and anchor locations: Horizontal: Figures B29, B30, 343, B44, 347, B48, 349, B50, 353 and 355 attached. l Vertical: 2.0 times attached figures B62 and C12. The following site specific response spectra are applicable to the main steam support and anchor loca-tions: l l i l l .-~.m._

1 Answer Ic: Horizontal: Figures B29, B30, B43, B44, B51, BS2, (Cont'd) B53 and B55 attached. Vertical: Same as RRS vertical spectra noted Jouve. Question 2: Provide a comparison of all the design basis response spectra used in the BOP piping analysis to the cor-responding site specific spectra. Answer 2: With the exception of piping analyzed by the NSSS supplier (multiplier of 2.0), all large bore piping identified as being required for safe shutdown was evaluated for SSE response using a multiplier of 1.875 times the OBE response (%% damping). Several sample plots.f 1.875 times the FSAR OBE response spectra (%% damping) VRS. The same site specific re-sponse spectra (2% damping) are attached (see sheets A-5 through A-14). In addition, input SSE response spectra for two selected piping systems are plotted against the corresponding site specific spectra and included herein (sheets A-15 through A-20). Note that envelope response spectra should be the compari-son basis for piping systems. Question 3: Provide a list of the seismic spectra applicable to each piping subsystem identified in the systems re. quired to achieve safe shutdown. Answer 3: A list of the input site specific response spectra VRS stress analysis subsystem is attached. 1 Question 4: Provide the piping modal frequencies of each piping subsystem ident!.ed in the systems required to l l achieve safe shutdown. Answer 4: The modal periods of each piping subsystem are attached, i i ~...

GENERAL @ ELECTRIC l NUCt. EAR ENERGY DIVISION oocuutNT No. 22A2662 arv.No. 3 SMECT NO. G Praquancy, CPS 100 33 m ___. 10 3.3 2.0 . 1.0 r-- r-. tm :_r + .--...._~a__.. -+-: 3.0 7 M_ EG < 1/2% DAMPUiG - --f.. _3==m=k _.t==:-. _. - L, :.. _ _..t -.- m. _ _ EE.iEE:.:1 9._2. M *J.__.:==: g g I% DAMP UG , _.._,.-- e~. ; ..g Li ~ 2% DAF2ING ,- H.. , _ L .... a-... p b ~_.,.4.. _1 5: DAMFDiG t-- -= ,L 1.0 e _._f. o. g f.4 g - .-.n -- p-y s a --+ - - ! p.- Hw: r = ___ _.. -_-. _ _.--d :. =_ p- :_r_-. : -- r - r :. ; - - J :._--. - - + m ::=.. =. __ - - _ _ ~ _ _.. - . = = _.-t h\\ _ :. -r __----e m

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2.0 Paciod, See h STF.AM PI? DIG FIGURE E STEAM PIPING Hopi OW AL M9'deM E S?EC*EA Half saf e shutdown earthquake, Concrete Conta.incent Shield 8'-0" above RPV invert, North-South component. A-l Extracted from $ argent & Lundy, Job 3988, June 27,1972 s~ ~~

GENERAL @ ELECTRIC G NUCLEAR ENERGY DIVISION DOCUMENT NO. 22A2662 "EV No 3 sacrT No. 43 Frequency, CPS 120. 33 10 3.3 ,1 0 2.0 I J l! t i I f i 3 .f 4 3 6

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u. x _w ,f H EQUIP,DAMPIN " % y Q u s xx J ? 3 ._.___._l_____-- u< + A- - - .03 _ - - - - - + - - - - . i__.- I. i 1 1 .f I I v t t I I .01 ~- .01 .03 .1 .3 1.0 2.0 Period, Sec - bD w_/ FIGURE E 3 TEAM AND RECIRCULATION PIPING VERTICAL RESPCNSE SPECTRA Half safe shutdown earthquake, Reactor Pressure Vessel Base, extracted from Sargent & Lundy, Job 3988, Aug. 25,1971 f l A-E l i r

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GENERAL h ELECTRIC 22A2651 sw. wo. 66 NUCt. EAR ENERGY DIVISION R E V. 2 Frequency, CPS 100 33 10 3.3 ~ 1.0 5 ..--}.l.3 [ _.ld f il t l

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j i . 0 L.- - _ _ _. _ l ,o1 03 .1 .3 1.0 2.0 t Period, Sec l FIGURE El-1 RECIRCULATION PIPING HORIZONTAL RESPONSE SPECTRA ' Half safe shutdca earthquakeTTop of Reactor ~ Pedestal, envelope of East-West and North-South h Extracted from Sargent & Lundy, Job 3988, April 18,1972 A-9 y .g-w-- -p-, - + - - -.. - ,pr-m-m--- --v.-

Ennico Fuaw - Unir 2 SHT. OF ~ Stl Swc. Reassrssurwr u N AMt.. t Det: Q A N A'.Y t f D Ntw sst Fateutwcy, CPS 50 33 2o 10 5.o 3.1 2.o io .5 ap ieie iii i i i i e a i iiii oi> iiii iiii e i e i i iiii > i i ' Zo o sa ~ ~ ita e.o _ 40 4.0,' ~ 6b 6h _;e O $0 50 O Ab ~ ~ ~ ~ 3b 3b ~ ~ 2.0 24 4 F ii\\ 3 N - i.5 Lg 2 \\ D j e.* 5 _ '4 t h [4 j 4 9 // %t 6 6 r yj x, A .5 3 / A. i g f 4 W 4 ~~ \\ / \\ f j [ ,/ i 3 3 d .3 / q s a / y xx \\\\ J5 s _ 4 \\ ~ ~ ~ '8 w .M ~.06 ~ ~ ~ A6 .06 ",og 4 ~ ns e i niin e e i e i it is i,i. .,e e i,si i i e i ii ni 64 43 .0 +.05. 06 46.I .L5 . ?. 3 .+ .5 6 8 to 1.5 2.0 Pe.nic o, S c c.

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i e 9 c.: SX-o.8 SARGENT 5 LUtOY C4/30/51 SX-05 6139-33 tX -03;C 5 J, 7 _ A952.CS F7.FC95C65510 FrRK ". 97.3 4.e. ..:.**'.t a c. 9 33 pre.c. 2 C.S 6M.72*.N."'.'4 '. cre.". D."'/. ~ 3 a *.*..":~. r, _' ~ -n c e.g m. ~.o. g..-,... . 3.A, pc.gwg .n w. M w,,t a :.'. ') 1 . C47 35, ^ J' .s.n ... u,s u.a ., ~ 4, 6' %'r :i. t . s..-. 7 . C 1, D. 3 3 .wc.:v 9 .013.:4 9, n 's ). n t. . s v,.,3 ., o .-:e 1 .sC. q a 1 c '4 Ca.?9 a3 te . Cvu: ~ . C.n;.,ic. ) 16 . t.. e.. 3 .s..,.. 19 .C M t 20 . CC.tT 9 .CC216 2,1 . Ar -r 2 ) 4 I. c: .vc 24 .CO.C0 a 45 .CC_m, 26 .00300 .00292 . ~ 3 CC, :"'- p .CsetO . i. ' - 40 .00267 CPU TPE TILL END CF EIGEN '/ alt.E FRCELEM '4A3.25 P~:i3 s'~

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..~ E' ? SX-DO SARGOiT & LUtOY 04/30/81 SX-10 6130-38 E.0 -c z.:,6 3 L.i i L_ s A974C8 PIP 095065510 FER513933 A S CI.50 P.S 2 a r 6139-38 SX-10 MCOAL PEP.ICOS ORTEG3C7" " O ^ ' :- ' ~ - nC'-' MCOE (SEC) Mc;d 9 :d; ' c,-.3, a 4 . v .05ho 4 .C5725 5 .0 5 1 6 .CIJ4 7 0-no '.0:: :u 7 3 02 .2 9 20 03613 ~ 11 .05421 ~' if .C2 ?';0 is .02i'io 14 .C2659 2 ^t* 3 t' S,

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~~ SAGGENT C tbM3Y 05/02/31 5X-13 6139-33 EMD 03007] 5; a i u; ~ "EISMIC R$,- 420TH PIP 095065510 FER613933 a 6139-38 SY-13 M00L DERI005 CETF: _"'t:? ' *0 .s FCCE (SEC) r :: i :Ti ' ~ .24394 2 .19720 3 .15022 4 .13173 5 09244 6 09092 7 .07212 8 06s?1 9 .05/33 10 .05524 .?- 11 .050$0 .1 12 .04952 13 .04576 14 04204 15 04152 j 16 .04052 17 .04000 .i~ -:~ ^ l 18 .03633 ~ 19 .03518 20 03367 .J.~-:- 21 .02943 22 02702 23 .02505 ~ ~ 24 02501 25 02345 26 .02236 . ~ _ ' 27 .02190 2? .02067

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'th ~ f f. s$ ,fgk.-fy'.1.-J .f. : ~.m r it.g5. x:q y. j ~ w_ ;. n,. , *[.% a , +.. EECW-16 SARGENT & LUNDY 04/25/81 EECW-?.5 6139-38 EMO-030057 S/R c# ?iL. Hi r:0 288C8 PIP 095065510 FER613938

  • SEISMIC R5
  • C3E5ZI2MI't't3

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  • 6-06 20 011 6

. 67 -c7 -21 .J.009 6 l r22 5 '-~ ; 'i.00970 .52T-07 .00949 .596-07 '23 24 00935 .532-07 .2n3-07 25 008 1 261 - l'~.007 6 .373-07 1 222-CF I 27 00741 .J47-C7 28 00736 .5E2-07 29 .C0729 . 30 < 00682 .671-07 s CPU TIME TILL END GF. EIGEN VALUE PROBLEM WAS.31 MINS e. ~ g P /

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  • SEIS.9:C 13 *

!? ~ 0 Mi 1 e~--, =~1.-- 61??-35 FERM -2 EEC*d-22 EMC 03CC50 S.?. 0:iEM CRTHCG0hALI7Y AND h:~?.?C : ~I"i T.-E ? MODAL PERIODS MODE SHAPES CF.AX :IV:3T!;:!53 4 MCOE (SEC) 571-07 ! f 1 .15473 596-07 2 .06611 671-C7 } 3 .C4!!4 .224-07 i 1 4 .04811 .522-07 t 5 .03964 .522-C7 I 6 .C3745 447-07 i 7 .03:33 .522-07 3 .03157 .671-07 9 .02717 .522-07 10 .02393 .522-07 1 11 .01933 447-07 12 . 01713 .671-07 13 .01457 s, .522-C7 14 .01450 47-07 15 .01261 .596-G7 16 .011?2 .671-07 17 01170 .596-07 13 .01036 .522-17 19 .01056 20 .01015 .5?2j! '-5 21 .C1CO3 .745-03 22 00950 .522-07 23 .00902 .596-07 24 00821 .596-07 25 .00774 7.95 SEC CPU TIME FOR FINAL EIGENVALUE SOLUTICM WAS am m e.a*

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~ I MCCAL PERIOOS ORTH: :. i MCDE (SEC) c!' : i_:7 ~ ' 1 .16113 p .e:-- .x-s. '3 .C916 4 .07253 5 05?37 6 .C5'3) 7 .046:3 8 .C4232 9 .03435 t 10 .02959 11 .027C6 l 12 .02135 f ~ 13 .01770 ._:- ~ f 14 .01753 i 15 .01510 16 01413 17 .01294 la .012"4 19 .C1130 l 20 .01121 21 .01032 22 .01045 23 .00c97 ~ 24 .00950 ..;:!-07 .C0923 _ -;~ 25 26 .00768 .2 -:7 27 .00638. .277 ^7 28 .00609 } 29 .00596 .'47-;' l 30 .00554 .?T - 7 { CPU TIME TILL END GF EIGEN VALUE PROBLEM WAS 11 MII ,o**'- ,s

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l 7.3.9.1 DESIGN BAISIS.INFORMATION l I The design basis information for the instrumenta-tion and contirol of the RHRSW System, as required in Section 3 of IEEE 279-1971, is provided in , Subsection 7.1.2.1.19. 7.3.9.2 SYSTEM DESCRIPTION i The RHRSW System provides cooling water to remove heat from the RHR System. The RHRSW l System includes a closed cycle supply of water, pumps, and mechanical draft cooling towers to reject the heat to the environment. The system will operate with or without a loss of offsite l.k power. The REP.SW System is described in FSAR I Subsection 9.2.5 and the system diagram provided in FSAR Figure 9.2-4. The following discussion l provides additional information on the RHRSW instru-ment and control. ,m ,,n. o L l l ~ ~ l i

~: 7.3.9.2.1 POWER SOURCES Instruments and controls for the RHRSW System receive electrical power from the redundant 120 volt - 60 Hz Instrument Power Systems des-cribed in FSAR Subsection 8.3.1. Part of the control logic is de, powered by the Class lE de system described in 8.3.2. The pressure con-trol valves requiring pneumatic power receive plant instrument air as described in FSAR Sub-section 9.3.1. 7.3.9.2.2 EQUIPMENT DESIGN Each of the two separate, redundant RHRSW loops has electrically and physically separate controls and instruments. 7.3.9.2.3 INITIATION AND CONTROL CIRCUITS The RHRSW pumps, valves, and cooling tower fans are all initiated manually from the main con-trol room f 4 y mv

j 7.3.9.2.4 LOGIC The RHRSW is a manually initiated system, there-fore, this is no automatic initiation logic. ~ The RHRSW pumps automatically trip if they are operating and a LOCA signal is received, as indi-cated in FSAR Figure 7.3-8, Sheet 1. This trip is provided to allow the automatic loading of other ESF equipment on the emergency diesel genera-tors if a loss offsite power o.ccurs. The inter-1 lock can be bypassed by a keylock switch so that the pumps can be started if there is a long i term LOCA signal present. The cooling tower fan motors automatically load shed if a loss of off-site power occurs. The motors must be manually reset from the main control room before they will restart. I l I 7'.3.9.2.5 Testability Each RHRSW loop can be tested from the control room by starting the RHRSW pumps and/or cooling tower fans. Any necessary calibration of the j instrument and controls is accomplished at that i time. If an accident signal is received during l a test, the system pumps trip off and remain off l until the operator manually restarts.

h,-[ EF-2-FS AR M 77K e e s (,f, 6f.%-D Y19 7.4.2.4 Reactor Shutdown Cooling System Instrumentation and Control 7.4.2.4.1 Ccnformance to General Functional Requirements The design of the reactor shutdown cooling system instrumenta-tion and controls meets all the functional requirements of Subsection 7.1.2.1.27 as follows: 7.4.2.4.1.1 yalves Manual controls and position indicators are provided in the main control room. Interlocks are provided to prevent opening of the valves if shutdown conditions are not met. Interlocks are also provided to close the valves if an isolation signal 4-is present or if high reactor pressure exists. _g7 " 1. 7.4.2.4.1.2 Instrumentation Shutdown flow indicator is provided. Heat exchanger cooling water and service water temperatures are provided. Head spray flow indication is provided. 7.4.2.4.1.3 Annuncia tion The following annunciators are provided: a. Valve motor overload b. Heat exchanger service water outlet temperature high + c. Heat exchanger shutdown cooling wa ter high temp-erature d. Shutdown suction header high pressure e. Pump overload ~ f. Discharge header high pressure. 7.4.2.4.1.4 Pungs } Manual controls and stop and start indicators are provided in the main control room. Interlocks are provided to trip the pumps if the shutdown cooling valves are not properly set up, f 7,. 4. 2. 4. 2 Conformance to Specific Regulatory Requirements There are no specific regulatory requirements for the instru-mentation and controls of this system because this subsystem of the RHR system is used only to cool the reactor core for removal of decay heat with the reactor fully shut down and at approximately 50 psia. 9 7.4-22 O'I* h e w gg

/ INSERT 7. 4, '2,, 4, t. ) Redundant sensors (NillA & B) are provided for the RHR shut-down cooling pressure interlocks. These sensing loops are felt to meet or exceed the EICSB-3 Branch Technical Position. The interlocks are designed as part of the testability option and as such formal diversity of the sensors and trip units has not been provided. It is Edison's position that the accuracy,

  • reliability, and the inherent on-line status monitoring of the analog transmitter-trip unit design obviates the need for diverse instruments.

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  1. rWpr 7 6f2 -DVW The manual initiation of the system need not occur until the pres-sure of steam trapped between the isolation valves decreases to the vessel pressure and abnormally high radiation is present.

The pressure decay between the MSIVs due to MSIV leakage at an assumed rate of 11.5 scfh at 29 psid from the volume betwt.n MSIVs is such that the line pressure will exceed vessel pressure for more than one hour, after which the vessel and trapped line pres-21 sure will be about 35 psid. There would be no need to actuate any portion of the MSIVLCS when the pressure between valves is higher than containment pressure, because clean steam, trapped between the valves when they closed, would be leaking toward the containment. The instrumentation necessary for control and status indication of the MSIVLCS are classified as essential, and as such are designed in accordance with IEEE 344-1971, IEEE 279-1971, and IEEE 323-1971. They are qualified to function under seismic Class I and LOCA environmental loading conditions appropriate to their installation with the control circuits designed to satisfy the mechanical and electrical separation criteria. The system will detect high-steam line pressure and prevent system actuation. It will also detect high leakage and prevent release of leakage. 9A.4 DESIGN OF MAIN STEAM SYSTEM PIPING AND VALVES The main steam piping system, from the outboard MSIV to the appro-priate anchor positions of all branch lines downstream of the third MISV is seismically qualified. The main portion of the main steam system is located in the turbine building, which is seismically qualified to withstand the ef fects of an OBE or an SSE event (Reference 1). 't The main steam system has been seismically analyzed to ensure its integrity after either an OBE or an SSE event. The section of main steam piping analyzed begins at the anchor outside the primary containment and ends at the anchor in each of the branch 21 lines downstream of the third MSIV. The seismic analysis of this portion of the main steam piping and included valves is described in a report on the seismic qualification of the main steam piping (Reference 2). The report verifies that piping structural and pressure integrity will be maintained, and the included valves I will remain in the elastic stress range af ter either an OBE or an SSE event. The quality Group D+0A requirements are defined. in the following sections, QCk e 9A.4.1 Main Steam Lines The main steam lines from the outboard containmen : isolation valve up to the third MSIV and all the branch lines co nected to these ]) portions of the main steam lines up to the first valve in the branch line, are classified as Group O'. These sections of these ~ 9A-5 Amendment 21 - March 1979

n l EF-2-FSAR /d ~ \\c# and / pipes meet all pressure integrity requirements of Group / >\\ D 21 the following requirements. a. .Til longitudinal and circumferential butt weld joints are radiographed or equivalently tested ultrasonically. Where size or configuration do not permit effective volume examination, magnetic particle or liquid perie-trant examination is substituted. Examination proce-dures and acceptance standards are at least equivalent t those specified in ASME Section III for Class 2 21 components. b. All fillet and socket welds are examined by either magnetic particle or liquid penetrant methods. All structural attached welds to pressure retaining mate-rials are examined either by magnetic particle or liquid penetrant methods. Examination and acceptance 21 1 standards are at least equivalent to those specified I in ASME Section III for Class 2 components. c. All inspection records will be maintained for the life of the plant. These records include data pertaining g4jh;toqualificationof inspection personnel, examination procedures, and examination results. 9A.4.2 valves in Branch Lines Connected to Main Steam Lines The first valve in branch lines connected to the main steam lines between t e outboard containment isolation valve.and the tQird MSIV meet all the pressure-integrity requirements of Group and the following requirements. d) 21 a. Pressurc retaining components of all cast parts of ~~ valves of a size and configuration for which volumetric examination methods are effective, are radiographed. Ultrasonic examination to equivalent standards are used as an alternate to radiographic methods. Exam-ination procedures and acceptance standards are at least equivalent to those specified in ASME Section III for Class 2 components. b. All inspection records will be retained for the life of the plant. These records include data pertaining to the qualification of inspection personnel, examina-tion procedures, and examination results. N e b qs %) 'mendment 21 - March 1979 9A-6 A 3

~ EF-2-FS AR '~ ~ 10.3 MAIN STEAM SUPPLY SYSTEM 10.3.1 Design Bases 10.3.1.1 Safety Design Bases V"O O (,% To satisfy the safety design bases, the main steam lines from the reactor up to the tst!$serieXs$byvalves are designed according to the following piping classification which is in accordance with the ASME Boiler and Pressure Vessel Code: From the reactor to the drywell wall - ANSI B.31.7, a. Class A; Category I b. From drywell wall to the outer main steam isolation valve - Section III Class 1, Category I 4 ied ttod,. c. From outer isolation valve to the '97' valve - ANSI B.31.1.0, Category I, with volumetric examination. 10.3.1.2 Power Generation Design Bases The main steam supply system is designed to fulfill the following functions: To deliver steam from the NSSS up to the turbine-a. generator b. To provide steam for the reheater and the steam-jet air-ejectors To pro /ide steam for the RFP turbines during startup c. and low-load operations d. To provide steam for the turbine seal system and i i flange warming during startup To deliver excess steam produced in the NSSS to the e. 1 condenser during startup and transients whenever the steam used by the turbine is less than that produced ) by the NSSS. 10.3.2 Description The main steam supply system is shown in Figure 10.3-1. The main steam piping consists of four 24-inch lines from the outermost main steam line isolation valves to the 52-inch manifold, and then to the main turbine stop valves. The turbine stop ' valves and main steam line isolation valves may be tested g independently during plant operation. s 10.3-1 ~ M. _.. y.w - ,__._ ; _ ~ ~ ~ y -q .n_ u,,

EF-2-FS A. 9 'the main steam line pressure relief system, main steam line flow restrictors, and main steam line isclation valves are described i. Subsections 5.2.2, 5.5.4, and 5.5.5, respectively. 4 The design pressure-temperature rating of the main steam piping is 1250 psia /575'F, the same as the design pressure-temperature of the NSSS. The Category I design requirements are placed (1) o.p 2 the main steam piping f rom the reactor up to the 'tsEkl(Wgy W ed 60 valves, h @ ' h c W h > W and (2) g on all branch lines up to and indluding the first valve that is either normally closed or capable of automatic closure during all modes of normal NSSS operation. The main steam line is also anal-yzed for the dynamic loadings caused by fast closure of the tur-bine stop valves. A 52-inch manifold is installed ahead of the turbine stop valves. This provides a common point for the four steam lines from the reactor, the four steam lines to the turbine, the two bypass steam lines, the steam line to the reactor feed pump turbines, and plant auxiliaries. A drain line is connected to the low points of each main steam line, both inside the drywell and outside the containment. Both gets of drains are headered and connected by valving to permit steam line isolation and drainage to the main condenser hotwell. To permit draining the lines for maintenance, an additional drain is provided from the low point of the drains to the radwaste system. The drains inside and outside the containment are capable of ~ squalizing pressure across the main steam line isolation valves prior to restart following steam line isolation. Assuming all cteam line isolation valves are closed, and the steam lines out-side the drywell have been depressurized, the isolation valves outside the drywell are opened first. Then the drain lines are 1 used to warm up and pressurize the outside steam lines.

Finally, the main steam line isolation valves inside the drywell are opened.

10.3.3 Evaluation . The seismic and quality group requirements of all main steam lines and components are defined in Chapter 3. This design ensures con-l formance with the intent of NRC Regulatory Guide 1.26. I l 10.3.4 Insoection and Testing Requirements Inspection and testing are carried out in accordance with the requirements of NRC Regulatory Guide 1.68 and ANSI N18.7. The main steam line is hydrostatically tested to confirm leaktight-noss. All welding in the above steam line is 100 percent -volumetrically inspected. E) 10.3-2 Amendment 2 - January 1976

m

a g r Question O30 3.l, $$ a. Your proposed increase in leakage testing intervals for the main steam isolation valves (MSIVs) is unacceptable. Appendix J to 10 CFR Part 50 requires a test frequency of every refueling but never to exceed two years. It is our position that the test frequency meet the requirements of Appendix J to 10 CFR Part. 50 and be consistent with the BWR Standard Technical Specifications. Answer In order to comply with your position, the FSAR has ber. updated to include a change in the length of time between leal rate testing (attached). Our leak rate test will still require a maximum of 100 SCFH leakage from the entire system for each division. Individual valve leak rate tests will c9t be required. Therefore this test is still considered a " Type A Test." 1 -w .r- ,,-.,-,n-

EF-2-FS AR the preoperational testing and are used as base points for meas-urements obtained in subsequent operational tests. O The tests are conducted by pressurizing the system between the MSIVs and valve V10-2010 with air at pressures of 5, 10, 40, and 80 psig. Pressure decay is recorded as a function of time and the leak rate is computed with consideration of MSIV leakage, temperature correction, system volume measurement, relative humidity, and instrumentation tolerances. All valves are then exercised at their normal operation for 10 cycles and the leak tests are repeated. In no case shall the Division 1 or Division 2 system leakage exceed 100 scfh at a differential pressure of 80 psid. During plant operations the valves, piping, instrumentation, wir-ing and other components outside the steam tunnel can be inspected visually at any time. Components inside the tunnel can be in-spected only when it is open for access. Valves located in the tunnel can be exercised periodically during normal operation, pq Main steam system leak tests are conducted __ l-___ 2 = _ - -i= - A a . + _m g - 2.!> _l' 72 __ ' Leak tests are cbndbbted using Divisibn 1 abd S Al-Division 2 of the MSIVLCS at a differential pressure of 80 psid, 21 with an air temperature of 700F. The Division 1 or Division 1% E g[ 2 leakage is not to exceed 100 scfh when corrected for leakage controlling parameters. Mb The tests may be conducted during normal refueling periods or M unscheduled shutdowns. If the leakage were to be in excess of du= the Technical Specification limits the valves would be reworked

  • 4 and the system retested.

The Technical Specification limit for g the main steam system leakage (MSIVLCS Division 1 or Divicion 2) $g m is 100 scfh at 80 psid differential pressure and 700F when tested -) with air at a dew point of -200F or below. 9A.3.5 Instrumentation am l Sufficient instrumentation is provided in the main control room $) to allow the plant operator to assess the status of the MSIVLCS. ? g The open-close status of the eight inboard / outboard MSIVs, third E MSIVs, and all MSIVLCS solenoid-operated valves are provided by }^ l indicating lights. Pressure indication is provided for the volume a m bounded by the inboard / outboard MSIVs, the volume bounded by the p$ outboard MSIVs and the third MSIVs, and divisional control air header pressure. t Flow indication for each divisional MSIVLCS is provided. Low 1 differential pressure alarms are provided to alert the plant o operator of the possible failure of an inboard MSIV, outboard g MSIV, or third MSIV to be in the fully closed position. Low-m and high-pressure alarms are also provided to alert the plant operator that control-air header supply pressure is being lost, or that an injection valve is failing. e 9A-4 Amendment 21 - March 1979

Question 0 2a.).t, Afd-b. You have also proposed to increase the technical s_ ecification allowable leakage rate of the MSIVs to 100 SCFH. We agree that this increase is acceptable if you verify that.the other portions of your FSAR are revised to reflect this change. Specifically, Chapter 9A pressure decay assumptions for the line pressure between MSIVs assumes a leakage rate of 11.5 SCFH instead of 100 SCFH. Your basis for decay time for initiating the MSIV-LCS is dependent upon this leak rate. The time following a LOCA that the LCS is initiated should be calculated assuming the limit of 100 SCFH. Also, if necessary, revise any other areas of your FSAR, including accident analysis in Chapter 15 that are based upon an 11.5 SCFH leakage rate rather than 100 SCFH. Answer b. The FSAR has been changed to reflect 100 SCFH maximum leak rate of the l'SIV system (attached) and the conclusions have not chang ed. The assumed rate of 11.5 SCFH allowed the pressure trapped between valves to remain positive for more than one hour. This time period was very conservative and at 100 SCFH, the pressure trapped between valves is still more than one j hour. This will allow ample time for operator action to ensure i positive sealing and zero leakage. No other section is required to be revised to reflect this change. l l l l l [ l

EF-2-FS AR The manual initiation of the system need not occur until the pres- -}, sura of steam trapped between the isolation valves decreases to the vessel pressure and abnormally high radiation is present. The pressure decay beJween the MSIVs due to MSIV leakage at an assumed rate of "isefh e>eth=essak from the volume between MSIVs is such that the line pressure will exceed vessel pressure for more than one hour, after which the vessel and trapped line pres-21 sure will be about 35 psid. There would be no need to actuate any portion of the MSIVLCS when the pressure between valves is higher than containment pressure, because clean steam, trapped between the valves when they closed, would be leaking toward the containment. The instrumentation necessary for control and status indication of the MSIVLCS are classified as essential, and as such are designed in accordance with IEEE 344-1971, IEEE 279-1971, and IEEE 323-1971. They are qualified to function under seismic Class I and LOCA environmental loading conditions appropriate to their installation with the control circuits designed to satisfy the mechanical and electrical separation criteria. The system will detect high-steam line pressure and prevent system actuation. It will also detect high leakage and prevent release of leakage. 9A.4 DESIGN OF MAIN STEAM SYSTEM PIPING AND VALVES iT The main steam piping system, from the outboard MSIV to the appro-priate anchor positions of all branch lines downstream of the third MISV is seismically qualified. The main portion of the main steam system is located in the turbine building, which is seismically qualified to withstand the effects of an OBE or an 4,' SSE event (Reference 1). The main steam system has been seismically analyzed to ensure its integrity after eithar an OBE or an SSE event. The section of main steam piping analyzed begins at the anchor outside the primary containment and ends at the anchor in each of the branch 21 lines downstream of the third MSIV. The seismic analysis of this portion of the main steam piping and included valves is described in a report on the seismic qualification of the main steam piping (Reference 2). The report verifies that piping structural and pressure integrity will be maintained, and the included valves will remain in the elastic stress range after either an OBE or an SSE event. The quality Group D+QA requirements are defined in the following sections. t 9A.4.1 Main Steam Lines The main steam lines from the outboard containment isolation valve up to the third MSIV and all the branch lines connected to these .j}) portions of the main steam lines up to the first valve in the branch line, are classified as Group D+. These sections of these 9A-5 Amendment 21 - March 1979 ~

WWW9 SF2 - cNg SAVANNAH RIVER QUESTION NO. 11 Section 6.3 of the SRP requires that no operator actions should be required for the first 20 minutes after a LOCA. The appli-cant must describe all actions assumed prior to 20 minutes.

RESPONSE

ECCS ANALYSIS As indicated in Section 6.3.2.8, no operator actions are assumed for 10 minutes after a postulated LOCA. Of the five criteria specified in Section 50.46 and Appendix K to 10 CFR 50, the maximum peak cladding temperature, maximum cladding oxidation, maximum hydro-gen generation, and phenomena which might jeopardize maintaining coolable geometry, all occur before 10 minutes for the design-basis accident. t.ny break outside the primary containment in a line which connects directly to the reactor pressure vessel may need operator action. This is because there will be no high dr%well pressure signal to automatically signal the automatic depressurization system (ADS) to actuate. When the main steam line isolation valves (MSIV) close and the break becomes isolated or is too small to significantly depressurize the vessel, inven-tory makeup can only be supplied by the high pressure systems. For this reason the failure of the DC power source (for the division with the HPCI, 2 LPCI, and 1 core spray system) is the worst single failure. Given l

~ Page 3 Containment Analysis For the Mark I Containment Long Term Program Plant Unique Analysis, Appendix A of NUREG - 0661 limits the duration of generic SBA con-densation loads at 10 minutes into the accident by assuming manual operation of the Automatic Depressurization System. NUREG - 0661 allows that longer time periods maybe assumed for the SBA provided the eff ects are addressed in the PUA. The schedule for completion of the Fermi-2 LTP-PUA was provided in Edison letter EF2-53476, " Fermi-2 Containment Long Term Program - Plant Unique Analyais Schedule, " W. F. Colbert to R. L. Tedesco, June 8,1981. The limiting duration for the SBA condensation loads will be addressed i in the Fermi-2 LTP-PUA. i l l 4 F}}