ML20102C170

Forwards Status of Open & Confirmatory Items Identified in SER Sections 1.7 & 1.8.Resolutions to Listed SER Items Also Encl for Review & Approval.Info Will Be Incorporated in Amend 10 to FSAR

ML20102C170
Person / Time
Site:	Hope Creek
Issue date:	03/01/1985
From:	Douglas R, Mittl R Public Service Enterprise Group
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8503050330
Download: ML20102C170 (94)

Text

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Pubhc Serwce O PS Elecinc and Gas Cornpany 80 Park Plaza, Newark, NJ 07101/ 201430 8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L Mitti Genera! Manager Nuclear Assuranc.e and Regulation March 1, 1985 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Attention: Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

SAFETY EVALUATION REPORT OPEN AND CONFIRMATORY ITEM STATUS HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Attachment 1 is a current list which provides a status of the open and confirmatory items identified in Sections 1.7 and 1.8 of the Safety Evaluation Report (SER). Items iden-tified as " complete" are those for which PSE&G has provided responses and no confirmation of status has been received from the staff. We will consider these items closed unless notified otherwise. In order to permit timely resolution of items identified as " complete" which may not be resolved to the staff's satisfaction, please provide a specific description of the issue which remains to be resolved.

Enclosed for your review and approval (see Attachment 3) are the resolutions to the SER items listed in Attachment 2.

This information will be incorporated, as requ. ired, into Amendment 10 of the HCGS FSAR.

Should you have any questions or require any ad11tional information on these items, please contact ua.. .

Very tru,1y yours, i hu- g prism n8886 PDR i

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1. \

D Attachments ( l The Energy People

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Director of Nuclear Reactor Regulation 2 3/1/85 C D.-H. Wagner USNRC Licensing Project Manager (w/ attach.)

A. R. Blough USNRC Senior Resident Inspector (w/ attach.)

M P84 154/04 1/2 1

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Date: 3/1/85 F

ATTACHMENT 1 R. L. Mittl to Subject A. Schwencer -

Itra No. Status ltr. dated 01-1 Riverborne Missiles Partial Response 1/31/85 & 2/22/85 01-2 Equipment Qualification Partial Response 2/1/85, 2/20/85,

& 2/28/85 OI-3 Preservice Inspection Program Partial Response 2/14/85 OI-4 GDC 51 Compliance OPen OI-5 Solid-State Logic Modules NRC Action OI-6 -Postaccident Monitoring NRC Action Instrumentation OI-7 Minimum Separation Between Open Non-Class IE Conduit and 1

[ Class IE Cable Trays OI-8 Control of Heavy Loads Completed 1/18/85

! OI-9 Alternate and Safe Shutdown NRC Action 01-10 Delivery of Diesel Generator Closed Amendment 8 j Fuel Oil and Lube Oil 01-11 Filling of Key Management Open i

! Positions 01-12 Training Program Items (a) Initial Training Program Completed 1/7/85 (b) Requalification Training Completed 12/28/84

, Program (c) Replacement Training Completed 1/7/85 Program I

(d) TMI Issues I.A.2.1, Completed 1/7/85 I.A.3.1, and II.B.4 (e) Nonlicensed Training Completed 1/7/85 Program 4

01-13 Emergency Dose Assessment Closed 1/7/85 Computer Model 01-14 Procedures Generation Package Closed 1/28/85 01-15 Human Factors Engineering Open M P85 27/10 1-mr

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2 i

R. L. Mitti to A. Schwencer

-Itsm No. Subject Status ltr. dated

' C- 1 Feedwater Isolation Check Open -

valve Analysis

(-

C-2 . Plant-unique Analysis Report Completed 1/8/85-& 1/31/85 C-3 Inservice Testing of Pumps and 'Open Valves C-4 Fuel Assembly Accelerations Completed Amendment 8 C-5 Fuel Assembly Liftoff Completed Amendment 8 C-6 . Review of Stress Report Open C Use of Code Cases Completed 12/17/84 C-8 Reactor Vessel Studs and Fastners Completed 2/15/85 C-9 Containment Depressurization NRC Review Analysis C-10 Reactor Pressure Vessel Shield NRC Review Annulus Analysis-C-ll Drywell Head Region Pressure NRC Review Response Analysis e C-12 Drywell-to-Wetwell Vacuum Breaker NRC Review Loads C-13 Short-Term Feedwater System Open Analysis C-14 Loss-of-Coolant-Accident Analysis Completed 3/1/85 C-15 Balance-of-Plant Testability Completed Amendment.8

. Analysis C-16 Instrumentation Setpoints Completed 2/15/85 C-17 Isolation. Devices Open C-18 Regulatory Guide 1.75 NRC Review C-19 Reactor' Mode Switch _NRC Review C-20 . Engineered Safety Features Open Reset Controls M P85.27/10 2-mr 1

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3 R. L. Mittl to A. Schwencer Item No. Subject' Status ltr. dated C-21: High Pressure Coolant Injection Open Initiation

.C-22 IE Bulletin 79-27 Completed Amendment 8 C-23 Bypassed and Inoperable Status NRC Review Indication C-24 Logic for Low Pressure Coolant Open Injection Interlock Circuitry C-25 End-of-Cycle Recirculation Pump Completed 3/1/85 Trip C-26 Multiple Control System Failures NRC Review C-27 Relief Function of Safety / Relief Completed 2/15/85 Valves C-28 Main Steam Tunnel Flooding Open Analysis C-29 Cable Tray Separation Testing Open C-30 Use of Inverter as Isolation Open Device C-31 Core Damage Estimate Procedure Open C-32 Continuous Airborne Particulate Open Monitors C-33 Qualifications o'f Senior Radiation Open Protection Engineer C-34 Onsite Instrument Information Open C-35 Airborne Iodine Concentration Open Instruments C-36 Emergency Plan Items Partial Response 11/9/84, 1/16/85, &

2/7/85 C-37 TMI Item II.K.3.18 Partial Response 3/1/85 M P85.27/10 3-mr-

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,. ATTACHMENT 2 I

5 ITEM NO. SER SECTION SUBJECT i C-14 6.3.5 and 15.9.3 Loss-of-Coolant-Accident '

Analysis C-25 7.6.2.4 End-of-Cycle Recirculation Pump Trip i

C-37 15.9.3 TMI Item II.K.3.18 i

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ATTACHMENT 3

SER Confirmatory Item No. 14 (SER Section 6.3.5 and 15.9.3) 6 Loss-of-Coolant-Accident Analysis The LOCA analysis reported in the FSAR were for a lead plant representative of Hope Creek. The applicant has committed to supply plant-specific LOCA analyses in a later amendment to the FSAR before fuel loading. The NRC staff will report

.the results of its review of the plant-specific analyses in a supplement to this SER. This is a confirmatory item.

Response

HCGS FSAR Sections 1.3, 1.10, 1.14, 6.2, 6.3, and 15.6 and i . Questions Responses 440.0, 440.27, and 440.28 have been ,

revised to reflect the results of the HCGS plant-specific ECCS analysis.

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'JS:vw MP85.48 03.2-vw

FGS F3&B 8/8J TABLE 1.3-1 Page 1 of 6 C05PARISCW OF 30 CLEAR STEAR SUPPLI SYSTEM DESIGN CM&BACTERISTICStaa Mope Creek Satch 1 Limerick Susquehamaa sua 4/5 BNR 4 Bus 4/5 ama 4 251-764 2]8-560 251-764 258-764 IhSK541_ABd Hydrag h gg31gg (Sectios 4.4)

Rated power, But 3293 2436 3293 J29J Design power, Ett (ECCS design basis) $wYr.3 N O 2550 3435 34J9 3 Steam flow rate, Ib/h 14.156 E6 30.03 36 34.156 E6 83.48 26 Core coolaat flow rate, lb/h , 100.0 E6 78.5 E6 100.0 56 100.0 16 Feedwater flow rate, Ib/h 14.117 E6 10.445 E6 14.317 E6 13.574 A6 System pressure, nominal is steam done, psia 1920 3020 1020 1020 Average power density, ku/ liter 48.7 51.2 48.7 48.7 saximaa linear heat generation rate, AN/ft 13.4 13.4 13.4 13.4 Average linear heat generation rate, ku/ft 5.34 7.11 5.3 5.34 sazians heat fins, sta/h-fta 361,600 428,300 361,600 361,000 Average heat flux, Sta/h-fta 144,100 164,700 143,700 144,100 saximas som temperature, *F 3412 4380 3435 3330 Average volaaetric fuel temperature, 'F 2149 2781 2130 2130 Average cladding surface temperature, *F 566 558 566 558 51minua critical power ratio 1.20 (*3 1.24 1. 2J Coolant enthalpy at core inlet, sta/lb 526.1 526.2 526.1 521.8 Core maxiana exit voids within assemblies 77.1 79 77.1 76.00 Core average exit quality, 5 steam 14.1 12.7 14.3 13.2 Feedwater temperature, *F 419.9 387.4 420 383

$[A / 7B/7? C ~/W Aseadment 1

i HCGS FSAR 10/84 i

performing the necessary work and submitted this information for staff review and approval.

l

Response

l General Electric provided information concerning the NRC's small-break-model concerns in a meeting between GE and the NRC staff 1

held on June 18, 1981 and subsequent documentation included in a i letter from R.H. Bucholz (GE) to D.G. Eisenhut (NRC) dated June 26, 1981. Based on its review of this information, the NRC staff has prepared a draft safety evaluation report (SER) that

concludes the test data, comparisons, and other information i submitted by GE acceptably demonstrate that the existing GE

! small-break model is in compliance with 10 CFR 50, Appendix K and, therefore, no model changes are required. "'- - """ "

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vi : Of the dr:ft S5S ::10: :ny furth r ::n;;rn:r',_

th:y eill be r:0 lved prier to the initiat!On Of the MCCS _4 __ t.

pecific ECCC :n !7:i: in 1 t: ! ?* d . e

• II.K.3.31 PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR 50.46 Position i

t Plant-specific calculations using NRC-approved models for small-1 break loss-of-coolant accidents as described in II.K.3 item 30 to show compliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees.

i Calculations to be submitted by January 1, 1983 or 1 year after

. staff approval of loss-of-coolant accident analysis models, i

whichever is later (required only if model changes have'been

! made).

Response

. Small-break LOCA calculations are described in Section 6.3.3.7, l and the results are summarized in Table 6.3-4. The references in

Section 6.3.6 describe the currently approved Appendix K methodology used. Compliance with 10 CFR 50.46 has been

previously established by the NRC. No model changes are necessary (see response to item II.K.3.30).

{

1.10-86 Amendment 8 l

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HCGS FSAR 8/83 1.14.1.26.2 Response This issue is not applicable to the HCGS because it does not have a HPCS.

1.14.1.27 Adequate Core Coolina Maintained with LPCI' Diversion, LRG 1/RSB-18 1.14.1.27.1 Issue l l

The NRC staff asked for a demonstration that adequate core cooling would be maintained if the flow of the low-pressure coolant injection were diverted to the wetwell and drywell sprays and to suppression pool cooling.

1.14.1.27.2 Response l l

i This situation is addressed in Section 6,3. Sufficient margin exists in the peak cladding temperature to accommodate the diversion of low-pressure coolant injection at 600 seconds into the transient. This demonstrates adequate core cooling. Turtt.cr G =

confir-etien vill be previded in the plant-unique ECCS ::: lyric e-th:t will bc ::=pleted in July 1925ve_.

l 1.14.1.28 Temperature Drop with Feedwater Heater Failure, LRG I/RSB-19 1.14.1.28.1 Issue 1

i The analysis of the feedwater heater failure event is based on a temperature drop no greater than 1000F. However, an actual failure demonstrated a 1500F drop. The NRC staff has requested a justification for the smaller temperature drop or a reanalysis with a justified temperature decrease.

1.14.1.28.2 Response The design specification for the feedwater heating system requires that the maximum temperature decrease due to a single failure be no greater than 1000F. Sufficient analyses have been ScR / rCo1 C -/9' 1.14-23 Amendment 1 l

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hCGS FSAR 8/83 1.14.1.107.2 Response i

l See response to LRG Issue No. 106, Section 1.14.106.

1.14.1.108 Nonconservatism in the Models For Fuel Claddino '

Swellina and Rupture LRG I/CPB-2 and LRG II/1-CPB 1.14.1.108.1 Issue The procedures proposed in NUREG-0630 introduce additional conservatism in the models for fuel cladding swelling and rupture during a loss-of-coolant accident. To assure the degree of swelling and incidence of rupture are not underestimated as required by Appendix K of 10 CFR 50.46, supplemental calculations to the current ECCS analyses should be performed. If the swelling is underestimated, the bundle cooling may be overestimated, and the peak cladding temperature may be ,

nonconservative.

1.14.1.108.2 Response

)

Abb INSC KT b The current under t:nding with the NRC ct:ff is thet the ECCS-nalyze: h.sve adequate evera.11 :::: rv: tic :lthough they ::y underecti::te th Offect: Of cladding crelling :nd rupture. When the MCCS-unique ECCS ::1culation: ::: prep::cd, in July I 1^05, the-eurve fer perfor:ti n et+00: veree: t::?er:ture will 5 -

Odified fer t: per:ture: beler 1500aF, :nd the then current

d 1 technol gy will be utili: d.

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f 1.14.1.109 Fuel Rod Claddino Balloonino and Rupture l 1.14.1.109.1 Issue l The procedures proposed in NUREG-0630 introduce additional conservatism in the models for fuel cladding swelling and rupture I during a loss-of-coolant accident. To assure the degree of l swelling and incidence of rupture are not underestimated as I required by Appendix K of 10 CFR 5046, supplemental calculations )

to the current ECCS analyses should be performed. If the swelling is underestimated, the bundle cooling may be overestimated, and the peak cladding temperature may be nonconservative.

gg gg g _fy 1.14-94 Amendment 1

INseAT &

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Mc9s- smgue ECCS e.lsJa. fins sus pnp J ufII.5"if ^ /*'" P Sa

• ined.1 anoJNii.d for b y.calaa.s dd' i 'k a b lo co f" r The Nec M.# bJ 4. m.J.I .uepis&I. win 7c4 Y. 4w, ceif.na. in Nuese,-os,ao as emdsseed bg s suppl s m }u ule.h e.ustusluk ud Mun, p.,+

4 g gA 41 model (see Refe nce in Sec.fsn kN. tor.2.2 I.h ut...-

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l l.14.106 2.1 Refe.n.nce ksk Arm H. Seend (MRc.) is s.ca. sacw,,i(a),

"SufftJ e bg Acaspbce o9 Licemi Top,W y AlEDE -Sota-P,' MFM 067-83. , Mg it , Mar. .

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HCGS FSAR 10/84 CHAPTER 6 ENGINEERED SAFETY FEATURES

, FIGURES (cont)

J Fiaure Title 6.3-11 Head vs Low Pressure Coolant Injection Flow Used in LOCA Analyses 4

1 6.3-12 Process Diagram, Residual Heat Removal System la 6.3-13 RHR (LPCI) Pump Characteristics m o.x lin u.iri

)- 6.3-14 Peak Cladding Temperature and Mtn4 mum Local

{ Oxidation vs Break Area j

l 6.3-15 Normalized Core Power vs Time

, 6.3-16 Core Average Pressure vs. Time After Break i (DBA, Recirculation Suction Break, Failure of Channel A DC Source) 1 l 6.3-17 Normalized Core Average Inlet Flow vs Time After Break (DBA, Recirculation Suction Break, Failure

of Channel A DC Source)

6.3-18 Core Inlet Enthalpy vs. Time After Break (DBA, Recirculation Suction Break, Failure of

] Channel A DC Source) l l

6.3-19 Minimum Critical Power Ratio vs. Time After

! Break (DBA, Recirculation Suction Break, j Failure of Channel A DC Source) i 6.3-20 Water Level Inside Shroud vs Time After Break

, (DBA, Recirculation Suction Break, Failure,

Channel A DC Source) 6.3-21 Reactor Vessel' Pressure vs. Time After Break (DBA, Recirculation Suction Break, Failure of

, Channel A DC Source) i

! 6.3-22 Fuel Rod Convective Heat Transfer Coeffi-cient vs. Time After Break (Large Break

. Model) (DBA, Recirculation Suction

! Break, Failure of Channel A DC Source) i

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6-xv Amendment 8 l l SCB / 7T/71 6 -/Y l l

- . - . . -- . . _ . -.-,--- ..--.---.- -- -.--. D

HCGS FSAR 10/84 CHAPTER 6 ENGINEERED SAFETY FEATURES FIGURES (cont)

Fiaure Title (0.09 ft2 Recirculation Suction Break, Failure of Channel A DC Source) 6.3-42 Water Level Inside Shroud vs. Time After Break (Small Break Model)

<>. fta Recirculation Suction Break, Failure of Channel A DC Source) 6.3-43 Reactor Vessel Pressure vs. Time Afte Break (Small Break Model) ( fta o.

Recirculation Suction Break, Failure of Channel A DC Source) 6.3-44 Fuel Rod Convective Heat Transfer Coefficient vs. Time After Break (Small Break Model) ( Eh-2"I t 2 Recir- CV culation Suction Break, Failure of Channel A DC Source) 6.3-45 Peak Cladding Temperature vs. Time After Break (Small Break Model) ( W fta Recirculation Suction Break, Failure of Q

Channel A DC Source) .i '

6.3-46 Water Level Inside Shroud vs. Time After Break (Small Break Model) (Maximum Core Spray Line Break, Failure of1 Channel A DC Source) 6.3-47 Reactor Vessel Pressure vs. Time After Break (Small Break Model) (Maximum Core Spray Line Break, Failure of Channel A DC Source) 6.3-48 Fuel Rod Convective Heat Transfer Coefficient vs. Time After Break (Small Break Model) (Maximum Core Spray Line Break, Failure of Channel A DC Source) 6.3-49 Peak Cladding Temperature vs. Time After Break (Small Break Model) (Maximum Core Spray

s c R / rc /n c -/Y 6-xviii Amendment 8

s HCGS FSAR 10/84 i

4 CHAPTER 6 ENGINEERED SAFETY FEATURES i  !

FIGURES (cont) l t

l Fiaure Title h Line Break, Failure of Channel A DC 4

Source) i I

i 6.3-50 Water Level Inside Shroud vs. Time After  !

Break (Small Break Model) (Maximum Feedwater Line l i Break, Failure of Channel A DC Source) '

i 6.3-51 Reactor Vessel Pressure vs. Time After j

Break (Small Break Model) (Maximum Feedwater Line Break, Failure of Channel A DC Source) 6.3-52 Fuel Rod Convective Heat Transfer Coef-ficient vs. Time After Break (Small Break Model) (Maximum Feedwater Line Break, Failure of

Channel A DC Source) 6.3-53 Peak Cladding Temperature vs. Time After Break (Small Break Model) (Maximum Feedwater Line Break, Failure of Channel A DC Source) -

6.3-54 Water Level Inside Shroud vs. Time After Break (Maximum Main Steam Line Inside Pri ;ry "

Containment, Failure'of Channel A DC Source) 6.3-55 Reactor Vessel Pressure vs. Time After Break (Maximum Main Steam Line Break Inside ua Prirrr-; Containment, Failure of Channel A DC Source)

Abb zns CAL T i;> 4 6.3-56 4'F Water Level Inside Shroud vs. Time After Break (Maximum Main Steam Line Break Outside 4t Pri ry Containment, Failure of Channel A DC Source) 6.3-pf er? Reactor Vessel Pressure vs. Time After Break (Maximum Main Steam Line Break Outside

1. - Pri;;ry Containment, Failure of Channel A DC Source) 6.3-5F 60 Fuel Rod Convective Heat Transfer Coefficient vs. Time After Break (Small Break Model) ain Maximwn 6-xix Amendment 8

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i' n o e r-d /?

4.3-S4 fue/ Rod Con vecdive //ea d Tra nsfe r coe fficient vs. Time /?fder Br ea k

( Max imum Main Sfeam Lin e St-eak

.rns,'de

• Con lainm en d, /~a < /ar e o[

c hai>n e/ /? b e Source )

lo 3 -5 7 Peak c./a.olc/;ng 7 amperatar e vs. m m e.

. /?f der Break C rnaximum (nain sfe a m Line Break .Z n s i de Contain m en d,

/~ai /ut-e oS C.hanne/ .4 hc so u-r e e )

S C R. ITCm C. - 1 4

A - . _ . _a 4

• HCGS FSAR 10/84 CHAPTER 6 i

ENGINEERED SAFETY FEATURES FIGURES (cont)

Figure Title Steam Line Break Outside "ri;;;i # Containment, Failure of Channel A DC Source)

6. 3-Jf 6/ Peak Cladding Temperature vs. Time After Break (Small Break Model) (Maximum Main Steam Line Break Outside Containment, Failure of Channel A DC Source) 6.3-66 G4 Total Time Highest Powered Node Remains Uncovered

_ vs Break Area (Failure of Channel A DC Source) 6.4-1 Control Room Arrangement 6.4-2 Plant Layout with Respect to Control Room Intake 6.7-1 Main Steam Isolation Valve Sealing System, P&ID 6A-1 Model Schematic for Inadvertent Spray Actuation 6A-2 Thermal Heat Removal Efficiency of Containment Atmosphere Spray 6A-3 Containment Pressure Response - Inadvertent Spray Actuation - 2 Spray Loops, and 1 PV Fails 6A-4 Containment Temperature Response - Inadvertent Spray Actuation - 2 Spray Loops, and 1 PV Fails 6A-5 Differential Pressure Between Drywell and Suppression Chamber - Inadvertent Spray Actuation -

2 Spray Loops, and 1 VB Fails

6A-6 Containment Temperature Response - Inadvertent Spray Actuation - 2 Spray Loops, and 1 VB Fails 6B-1 Flow Diverter 6B-2 Reactor Shield Annulus Arrangement 4 6B-3a Schematic of the RPV Shield Annulus Model i

6-xx Amendment 8

.SCR / Tern e - / 9'

HCGS FSAR 8/83 conformance to Criterion 4 is demonstrated by conformance to Criteria 1 and 2 ,

e. Criterion 5, Long-Term Cooling "After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core." Conformance to Criterion 5 is demonstrated generically for General Electric BWRs in Section III.A of Reference 6.3-1.

Briefly summarized, the core remains covered to at least the jet pump suction elevation, and the uncovered region is cooled by spray cooling and/or by steam generated in the covered part of the core.

6.3.3.3 Sincle Failure Considerations The functional consequences of single failures, including operator. errors that might cause any manually controlled, electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS, and the potential for submergence of valve motors in the ECCS, are discussed in '

Sections 6.3.1.1.2 and 6.3.1.1.4. The most severe single failures are identified in Table 6.3-6. Therefore, only these single failures are considered in the ECCS performance analyses.

For 1:rge bre:he, f ilure Of ene of the SDC: i, ir general, the meet cerere frilure. Fer 05:11 breake, ler: Of MPCI 10 the T0ct-c:ver: f:ilure.

Abb :Cn d e r d /1 6.3.3.4 System Performance Durino the Accident i

In general, the system response to an accident can be described as the following:

a. An initiation signal is received.

I

b. A small lag time (to open all valves and run the pumps  ;

up to rated speed) occurs. '

c. The ECCS flow enters the reactor vessel. l l

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i 6.3-32 Amendment 1 l 1

$[g jfC/n 0 ~lY

IstSEAT Tor both large and small breaks. failure of the channel A de source is the most severe failure.

1 l

k A single failure in the ADS (one ADS valve) has no effect in large breaks. Therefore, as a matter of l calculational convenience, it is assumed in all calcu-lations that one ADS valve fails to operate in addition to the identified single failure. This assumption re- ,

duces the number of calculations required in the perfora- i ance analysis and bounds the effects of one ADS valve 1 failure and the channel A de source failure by themselves. ' i The only effect of the assumed ADS valve failure on the calculations is a small increase (on the order of 100*F) in the calculated temperatures following small breaks. l l

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HCGS FSAR 8/83 Immediately following a LOCA, the RHR system is aligned to the LPCI mode.

6.3.3.6 Limits Ani ECCS System Parameters Refer to Sections A.6.3.3.6 through A.6.3.3.7.2 of Appendix A of I Reference 6.3-3. j Compliance with Regulatory Guide 1.47 is identified in Section 1.8.

6.3.3.7 ECCS Analyses for LOCA 6.3.3.7.1 LOCA Analysis Procedures and Input Variables Refer to Section A.6.3.3.7.1, of Appendix A, of Seference 6.3-3.

The significant input variables used by the LOCA codes are given in Table 6.3-2 and on Figure 6.3-15.

I 6.3.3.7.2 Accident Description Reference to a detailed description of the LOCA calculation is provided in Section A.6.3.3.7.2, of Appendix A, of Reference 6.3-3.

6.3.3.7.3 Break Spectrum Calculations The 2n:1yci reculte presented in thi Ocetion were obtained frem- --

typic:1 LOCA analysis, which is represcatativc of' this plant "---

size and product line. A plant-specific LOCA analy;i: eill be~ -

~ _.

submitted later ;; :n FCAn ;;cndmentat__

A complete spectrum of postulated break sizes and locations is considered in the evaluation of ECCS performance. For ease of reference, a summary of all figures and tables in Section 6.3.3 is shown in Table 6.3-4.

A summary of the results of the break spectrum calculations is shown in tabular form in Table 6.3-3 and graphically on 6.3-34 Amendment 1 SCR IT*C m c. -/ '/

l HCGS FSAR 8/83 Figure 6.3-14. Conformance to the acceptance criteria (peak cladding temperature 5 22000F, local oxidation 5 17%, and core-wide metal-water reaction 5 1%) is demonstrated. Details of i

calculations for specific breaks are included in subsequent paragraphs.

6.3.3.7.4 Large Recirculation Line Break Calculations The characteristics that determine which is the most limiting large break are:

i a. Thecalculatedtimeforrefloodingthethh'otnode

b. The calculated time for uncovering the hot node i c. The calculated time of boiling transition.

The calculated time of boiling transition increases with decreasing break size, since the time of uncovering of the jet j pump suction inlet, which leads to boiling transition, is determined primarily by the break size. The calculated time for uncovering the h'ot node also generally increases with decreasing break size, since it is determined primarily by the reactor coolant inventory lost during the blowdown.

The hot node reflooding time is determined by a number of interacting phenomena, such as depressurization rate, i countercurrent flow limiting, and a combination of available ECCS.

The period between the uncovering of the hot node and its reflooding is the period when the hot node has the lowest heat transfer. Hence, the break that results in the longest period during which the hot node remains uncovered results in the highest calculated peak cladding temperature. If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be limiting, as it would have an earlier boiling transition time (i.e., the larger break would have a more severe result from a blowdown heat transfer analysis).

6.3-35 Amendment 1

.5 C R l7'C m C -/Y

HCGS FSAR 8/83 6.2 Figure 6.3-h( shows the variation with break size of the calculated-time the hot node remains uncovered. Based on these calculations, the design basis accident (DBA) was determined to be the break that results in the highest calculated peak cladding temperature in the 1.0 ft2 to 4.1 ft2 region (the largest possible area of a recirculation system line break is 4.1 fta),

Confirmation that this is the most limiting break over the entire break spectrum is shown in Figure 6.3-14.

Important variables from the analysis of the DBA are shown on Figures 6.3-16 through 6.3-25. These variables are:

a. Core average pressure as a function of time I

b. Core flow as a function of time l

c. Core inlet enthalpy as a function of time l

d. Minimum critical power ratio as a function of time l

e. Water level as a function of time l

f. Pressure as a function of time l

g. Fuel rod convective heat transfer coefficient as a function of time l

h. Peak cladding temperature as a function of time l

i. Hot pin (the rod with the highest cladding temperature at a particular time) average fuel temperature as a function of time

j. Hot pin fuel internal pressure as a function of time l The maximum average planar linear heat generation rate (MAPLHGR),

maximum local oxidation, and peak cladding temperature as functions of exposure (from the analysis of the DBA), are shown in Table 6.3-5.

6.3-36 Amendment 1  !

ScR t re in C - / +' ,

, . I HCGS FSAR 8/83 6.3.3.7.6 Small Recirculation Line Break Calculations l

Important variables from the analysis of the small break yielding the highest cladding temperature are shown on Figures 6.3-38 through 6.3-41. These variables are:

a. Water level as a function of time l l

b. Pressure as a function of time l

c. Fuel rod convective heat transfer coefficient as a function of time

d. Peak cladding temperature as a function of time l The same variables resulting from the analysis of a less limiting small break are shown on Figures 6.3-42 through 6.3-45.

6.3.3.7.7 Calculations for Other Break Locations Reactor vessel water level and pressuref med fuel rod convective heat transfer coefficient and the peak cladding temperature are shown on Figures 6.3-46 through 6.3-49 for the core spray line break, med on Figures 6.3-50 through 6.3-53 for the feedwater line break, Figures 6.3-54 M f.3-55 chew the rc ct:r vcc:el-

.::tcr Icvc' cr.d pre :ure for )G"mlin steam line break inside the l pri=:ry containment. ,g , 77 and on An analysis was also done' for a main steam line break outside the-pri ry containment. Reactor vessel water level and pressure, fuel rod convective heat transfer coefficient and peak cladding temperature are shown on Figures 6.3-54' through 6.3-54'.

18 6/

6.3.3.8 LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is concluded that the ECCS will perform its function in an acceptable manner and meet all of the 10 CFR 50.46 acceptance criteria, given operation at or below the l

i 6.3-38 Amendment 1 SCR ITCD1 L*/Y

HCGS FSAR 8/83 automatically realign from system flow test modes to the emergency core cooling mode of operation following receipt of an automatic initiation signal. The core spray and LPCI systems begin injection into the reactor pressure vessel (RPV) when reactor vessel pressure decreases to system discharge shutoff pressure. HPCI injection begins as soon as the HPCI turbine-pump is up to speed. The injection valve is open, since the HPCI system is capable of injecting water at full flow into the RPV over a pressure range from 200 psig to reactor pressure specified in mode A of Figure 6.3-3.

6.

3.6 REFERENCES

6.3-1 General Electric, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDF-20566-P, November 1975.

f 6.3-2 H. M. Hirsch, Methods for Calculatino Safe Test Intervals and Allowable Repair Times for Encineered Safeauard Systems, NEDO-10739, General Electric, January 1973.

6.3-3 General Electric, " General Electric Standard Application for Reactor Fuel," including the

" United States Supplement," NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved revision).

i

c. - / 6.3-43 Amendment 1 5CA /7~C/n

G .

Reacdor lau) press wee

'AffroX ors is reacheol . Core spea.y valve s re e ei J t. HCGS FSAR prstsu re P ermis siVC-

+o oPEG, s c3nal TABLE 6.3-t OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEM FOR j DESIGN BASIS LOSS-OF-COOLANT ACCIDENT (8)  !

Time (s) Events 0 Design basis loss-of-coolant accident is assumed to start; offsite power is assumed to be lost.

Approx. O Drywellhighpressurekndreactorvessellowwater level (level 3) are reached. All SDGs are signaled to start, reactor scram is initiated, and HPCI, core spray, and LPCI receive the first signal to l start on drywell high pressure.

Approx. / Reactor vessel low-low water level (level 2) is

, reached. .HPQI receives the.second signal to start.

,$ MPc r inj echor? vc /ve. is sign o / eel to open Approx. / Reactor vessel low-low-low water level (level 1) is reached. The second signal to start LPCI and core spray is given. The auto-depressurization sequence begins./nSIV8 o r e. S 'S n o I e ci to c10se.

$ppeoX,If

$>0 All SDGs are ready to load. 0.2 ::70: inj::ti:n L

17; i: ci;n: led t: ;;: "-Energizing of the core y spray and RHR (LPCI) pump motors begins.

/9pproA g27 The HPCI injection valve is open and the pump is at design flow, which completes the HPCI startup.

The h core spray pumps are at rated flow and '

the3 injection valves are open, which completes the hresPr h core spray system startupg.

SeeFigure The core is effectively reflooded, assuming the 6.3-20 worst single failure; heatup is terminated.

>10 min The operator shifts to containment cooling.

(1) For the purpose of all but the next-to-the last entry on this table, all ECCS equipment is assumed to function as designed.

Performance analysis calculations consider the effects of single equipment failures (see Sections 6.3.2.5 and 6.3.3.3).

(2) No s ye J,'t i.s -/a. ken ;n the Dan i oen ana.tysi,s Fae EccS in E klo tion o n 1.he hig h city we il p nes s u.ne. s og n a.I.

APfro* 45 'rh e L pc r pu.mp s o.e e. o.L r a.+ed F l o v> a n d rh e.

inj e c.+io n va l v e.5 a re op ers w h i c b c.om pl eM cS

%c L Pc z s y stem s+=r+ wp .

Sl' R sw) C~'Y

HCGS FSAR 8/83 l TABLE 6.3-2 Page 1 of 3 SIGNIFICANT INPUT VARIABLES USED IN LOSS-OF-COOLANT ACCIDENT ANALYSIS Variable Value A. Plant Parameters 3430 Core thermal power 3*85 MWt l 37 Vessel steam output 14.e4 x 10* lbm/h Corresponding percent of rated 105%

steam flow Vessel steam dome pressure 1055 psia t%ima.m area o f r ee. ire u.Is+ ion lin e. break 4 1 f l' s B. Emergency Core Cooling System Parameters Low Pressure Coolant Injection System Vessel pressure at which $295 psid (vessel to flow may commence drywell)

Minimum rated flowy at ,

40,000 gpm, at vessel pressure 20 psid (vessel to drywell)

Initiating signals $24 Low water level, or 1.0 feet above top of

)a(ctive fuel High drywell pressure )(2.0psig Maximum allowable time delay 40 seconds from initiating signal to pumps at rated speed Injection valve fully open )(LOsecondsaftermaximum suction break Core Spray System Vessel pressure at which $289 psid (vessel to flow may commence drywell) i 6250 gpm, at 105 psid 1

Minimum rated flow, at vessel pressure (vessel to drywell) 3 Amendment 1 Sc 4 / 7c /n e -/ V l

• ' ' * -m e ------r- , - , _ _ , , , - , , . , , , _ , , , . _ _ , _ , , _ _ _ , _ , , _ _ _ _ _ ,

.. _. . . . = - _ _

HCGS FSAR TABLE 6.3-2 (cont) Page 2 of 3 Value Variable Initiating signals %

Low water level, or X1.0feetabovetopof active fuel High drywell pressure /2.0psig mi n i,n am Meenies allowed (runout) 4909 gpm flow per loop 7000 Maximum allowed delay time 27 seconds from initiating signal to P e c.s wr e. o. 4 pump at rated speed b YW A' p c,k ; nj ec+s'en MInjection valve fully open $27 seconds after maximum valve may open break or s a seconds a f ter pressare. perm es s I v e. a n g n o 1, which eve F (3 g reo ter, Combined HPCI/ Core Spray System Minimum flow rate 5600 gpm (independent of vessel pressure) r . core sPM 6250 gpm, at Minimum ratedjfloy 105 psid (vessel to pump available, at vessel pressure suction)

Initiating signals M 96 feet above top of

-1;0.3 Low water level, or active fuel High drywell pressure %2.0psig Maximum allowed delay time 27 seconds (c. ore spray 3V d* l from initiating signal to gs seconds (ype / sysfem) rated flow available and  ;

injection valve wide open 3000

mays *mam"i.._-"- HPCI flow rate 4000 gpm  ;

l injected through the core spray sparger Automatic Depressurization System -

Total number of relief 5 valves installed with ADS function SCR is c m c. - 14 l

1 HCGS FSAR 10/84 TABLE 6.3-2 (cont) Page 3 of 3 Variable Value Number of ADS valve used in analysis va

/Y 33 Totalminimumflowcapacity[ M x 106 lbm/h, at at a vessel pressure 1125 peit '---- ' '-

ru peic.. ; 011-A b s +i,ne r gnitiating signals M a) Low water level,0.nel .0 feet above top of active fuel high drywell pressure, and a g2.0 psig (t) .

signal that at least one .45 psig RHR pump or one core spray system is running ump discharge pressure Of' S4 A b) Low water level,0.nd p.0feetabovetopof active fuel b ~

high-drywell-pressure bypass timer timed out,and a g minutes s ig no.1 frorn in4idig signal that at least one 14 5 psig (.no t enod eled )

RHR pump or one core spray system is running ump discharge pressure ga j9b5 dirn er- lay time for all initiating

[ignalscompletedtothetime s [20 seconds valves are open Abb ~Z A>S LA'T 8 ->

C. Fuel Parameters grog g jl, Fuel type hit 9US corr, Fuel bundle geometry 8x8 La ttice C Number of fueled rods per bundle 62 Peak technical specification 13.4 kW/ft linear heat generation rate Initial minimum critical power 1. 2 ratio Design axial peaking factor 1.4 i ,i..di..g fe-. initi: ting sign;;; .;;,;,in ty,; s (1[) iTP.i; an:1 Aicried ci;

ng;.ie -

'9 C.R.

No e, r-c e{, }

taken in the b8A LocA a nalys s's far-Mendment 8 ggy ELC S s y2itsn in,* % fien o n t h e h e*1,h

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HCGS FSAR TABLE 6.3-3

SUMMARY

OF RESULTS OF LOCA ANALYSIS Break Size Peak Cladding Peak Local Location Temperature, Oxidation, Sinole Failure CF  %

S0%

A. 4.1 fta (DBA)/ M09(*) JJ / . f recirc suction /

Channel A de source IS99 B. 1.0 fta/ Large 47+F 1) <1 recirc suction / break Channel A de source methods sis 6 Small 1464-( a ) <j break methods

/I9V W&6(a) <1 C. 0.09 fta/

recirc suction /

Channel A de source (2) Core heatup model, CHASTE - large break methods.

(a) Non-DBA reflood - small break methods.

1 i

I l

4 5

ScR irem c-iY

HCGS FSAR TABLE 6.3-4 SUsetARY OF FIGURES AND TABLES IN SECTION 6.3 r F Large Break Transition Break small Breaks Other Breaks oW Break size DBA 1.0 fta 1.0 fta 0.09 fta JWf f ta Core spray Feedwater Main steam Main steam line line line line FSAR section 6.3.3.7.46.3.3.7.5 6.3.3.7.5 6.3.3.7.6 6.3.3.7.6 6.3.3.7.7 6.3.3.7.7 6.3.3.7.7 6.3.3.7.7 Remarks - Large break Small break Worst Addi- - -

Inside the Outside the methods methods small tional contain- contain-break small ment ment break variables- .

Core average pressure 6.3-16 6.3-26 - - - -

Core average inlet flow 6.3-17 6.3-27 - - - - - - -

Core inlet enthalpy 6.3-18 6.3-28 - - - - - - -

Minimum critical power ratio 6.3-19 6.3-29 - - - - - - -

Water level inside shroud 6.3-20 6.3-30 6.3-34 6.3-38 6.3-42 6.3-46 6.3-50 6.3-54 6. 3-W .57 maaotor vessel pressure 6.3-21 6.3-31 6.3-35 6.3-39 6.3-43 6.3-47 6.3-51 6.3-55 6.3-lW .T9 Fuel rod convective heat 6.3-22 6.3-32 6.3,36 6.3-40 6.3-44 6.3-48 6.3-52 g,3 . ft, 6.3-$r 60 tranafer coefficient Peak cladding temperature 6.3-23 6.3-33 6.3-37 6.3-41 6.3-45 6.3-49 6.3-53 la.3 5 7 6.3- W 6/

Bot pin average fuel 6.3-24 - - - - - - - -

temperature Hot pin fuel internal 6.3-25 - - - - - - - -

pressure Miscellaneous Tables and Fleures Input variables . Table 6.3-2 and Figure 6.3-15 Operational' sequence of ECCS for DBA Table 6.3-1 Peak cladding temperature, maximum local oxidation, and MAPLHGR versus exposure Table 6.3-5 summary of results of IDCA analysis Table 6.3-3 eingle failure evaluation Table 6.3-6 BCCS head versus flow curves Figures 6.3-4, 6.3-5, 6.3-9, and 6.3-11 Peak cladding temperature and maximum local oxidation versus break area Figure 6.3-14 Total time highest powered node remains uncovered versus break area Figure 6.3-fif 4.T e

S C R I TL m C -* */

HCGS FSAR TABLE 6.3-5 1

MAPLHGR, MAXIMUM LOCAL OXIDATION, AND PEAK CLADDING .l TEMPERATURE VERSUS EXPOSURE (1) G X3) l j

Average Planar Peak Cladding Exposure, MAPLHGR, Temperature, Oxidation mwd /t kW/ft 0F Fraction Fuel type 8CR183 200.0 12.0 1966 0. 97 i 1,000.0 12.1 1961 .0094 000.0 12.7 1981 0.0096 10, 0.0 12.8 1981 0.0094 15,00 0 12.9 2009 0.0128 20,000. 12.7 1997 0.0101 25,000.0 11.7 1883 0.0066 30,000.0 10.8 17 0.0042 ,

B. Fuel type 8CR233 200.0 1 9 1972 0.0098 1,000.0 12. 1961 0.0093 5,000.0 12.1 1937 0.0083 10,000.0 12. 1932 0.0080 15,000.0 1 .2 1957 0.0088 20,000.0 2.1 1960 0.0090 25,000.0 11.6 1909 0.0075 30,000.0 11.2 1855 0.0061 C. Fuel type 8CR71 200. 11.5 1878 0.0066 1,00 .0 11.4 1838 0.0056 5, 0.0 11.4 1806 0.0049 10 00.0 11.5 1792 0.0045

,000.0 11.5 1797 0.0046 20,000.0 11.0 1751 0.0039 25,000.0 10.4 1684 0.0029 30,000.0 9.7 1602 .0020 NEPl M e wi% i ns et-t O (1) The core-wide metal-water reaction has been calculated using method 1 described in Reference 6.3-1. The value is as i

follows:

O,/ C Core-wide metal-water reaction (%) = ScDF A d d In.se r-i e4 Sc& orem c -14

DNSECX (& lcf A) t .

A . Fo e t t y p e. P SC R B D 11 ,

200 11.5 1910 0.009 1,000 11.4 1872 0.008 5,000 11.4 1810 0.006 10,000 11.5 1794 0.006 15,000 11.5 1792 0.006 20,000 11.1 1747 0.005 25,000 10.4 1688 0.004 30',000 9.8 1621 0.003 35' ,000 9.1 1546 0.002 40,000 8.5 1468 0.001

- 45,000 7.8 1394 o.001 g , Fuel type P B c R BO 94 200 10.7 1912 0.009 1,000 11.0 1909 0.009 5,000 11.6 1879 0.008 10,000 11.9 1860 0.007 15,000 11.7 1820 0.006 20,000 11.3 1774 25,000 0.005 10.5 1694 0.004 30,000 9.8 1619 0.003 35,000 9.2 1547 40,000 0.002 8.5 1474 4.001

45,000 7.9 1407 0.001 C. Fuel h .a P BClt, B l 6 } ,

200 11.8 1990 0.012 1,000 11.8 1985 o .01 2 5,000 12.4 1994 0.011 10,000 12.8 1990 0.011 15,000 12.9 2015 0 .01 2 20,000 ,

12.8 2017 . o .01 2

25,000 12.2 1923 0.009 30,000 11.2 1788 0 006 j 35,000 10.6 1716 0.004 40,000 10.1 1658 0 003 45,000 9.4 1599 o.003 O

i SCR item c -' Y

-_,-. -- ....,--..e ,__,,,.,,.,___.,.,,_,,.,-.~,e,._,,__,,,_ _ ___,__,.,_,,,,m, _ _,. ,_ __ ,_ ,,,_.,,,,._ ,

msner @ '(sku+2 44

p. F_ o e t h, , e P s c R B2.4 8 200 12.1 2046 0.015 1,000 12.1 2037 o.014 i 5,000 12.3 1981 0.011 10,000 12.1 1949 0.010

' 15,000 12.1 1952 0 010 20,000 11.9 1941 0.010 25,000 11.2 1873 0.008 30,000 10.7 1790 0.006 35,000 10.0 1714 0.004 40,000 9.4 1650 0.003 45,000 8.7 1589 0.002 E. NI t ype. E 8 c it S 7_7 8 'r .

200 11.7 1960 0.011 1,000* 11.8 1957 c.011 5,000 12.4 1959 0.010 10,000 12.5 1951 0.010 15,000 12.4 1946 0.010 20,000 12.2 1936 0.009 25,000 11.5 1859 c.007 30,000 10.8 1779 0.006 35,000 10.2 1699 0.004 40,000 9.5 1633 0.003 45,000 8.1 1568 0.002 JW5&&T O

(2.) The analyses -- ' '"-- ^ were performed with the assumption that all lower tie plates are fully drilled.

Fenure 44-4)

(C) This analysis is valid for operation at all points on the power-flow map4 bounded by the most restrictive of the followings a) I,ess than the 1004-rated-power line b) I,ess than the APRM-rod-block line .

c) I,ess than the 1004-rated-core-flow line SC& ITE M C - l 'l

_= -. __ __ - . - - __ - .__

HCGS FSAR TABLE 6.3-6 SINGLE FAILURE EVALUATION The following table shows the single active failures considered in the ECCS performance evaluation.

Assumed Failure (8) Systems Remainina(a)

Y Channel A de source 1 core spray loop + 3 LPCI + g ADS SDG 1 core spray loop + HPCI + 3 LPCI +

Y5 MS LPCI injection valve 2 core spray loops + HPCI + 3 LPCI +

pg ADS HPCI 2 core spray loops + 4 LPCI + ADS J2 ADS valve 2 core spray loops + 4 LPCI + HPCI +

( 4 ADS

(*) Other postulated failures are not specifically considered, because they all result in at least as much ECCS capacity as one of the failures designated above.

< <a) Systems remaining, as identified in this table, are applicable to all non-ECCS line breaks. For a LOCA from an ECCS line break, the systems remaining are those listed, less the ECCS system in which the break is assumed.

4 SC A I TC /n C -/ C

- c_.-_ . . _ _ _ _ _ . . _ _ _ _ . _ . . _ _ _ _ . - . . . . _ . . , . - ,.

HCGS PSAR 10/84 TABLE 15.6-7 SEQUENCE OF EVENTS FOR A STEAM LI E BREAK OUTSIDE PRIMARY CONTAINMENT Approximate Time, s Event 0 Break of one main steam line outside primary containment AFBro^- 0.5 High steam line flow signal initiates closure of MSIVs becd

<l Reactorhcramg;^3YO

<5.5 MSIVs fully closed then SRVs open upon high vessel pressure. The j9pprot .5 0 valves open and close to maintain vessel pressure at approximately HOO psi.

w r / coo H66- ADS initiates on low water level, Ll' fcilarin;;

Appecx Vfo tim: del y: imp ^ erd by both th: ..OS "

ti=:r and-

/9 b.T ee high-drywell-pressure bypass timery s+o r+e.d .

A "; : 1 de; :;;uri;;; rapidly _

AffroX HM la/S Low-pressure ECCS systems begin injectionx Qith r-ea.c:to r h e.1 p a y1s a lit uncovered Apj>r /a 90 Core reflooded and clad temperature heatup terminated; no fuel rod failure.

hyprox. J 7 B es t and HPct would ha ve in >%kleof on /o to was er- -

le v e), L J U c!C e.onsidered un a vo o' lab /e o n c/ H P c / c..sousn eef l N d

sa. bled b y channe i n be power souree p,;iu ,e 3.

Aff ro k. 9 0 R e a cdor wo ter itVti a.bo v e c.o r e be a n.s To chrop s lo w l f d w e -t o th8 3888 o E 3 i " " ""0*S h 68 3 N' # 8 - R C A C YO"

( p ressu re t t m Aia 5 o+ *-PP r oxi m aie ly toco Psi.

,9pprox.970 19 tl A bs t im er 's + im e d elays wr e. co mp le t e d ; A b s valve a r e. uA u.a.+ ed ; rap;ol enep r esss e ; z. ant o n o G < e s s e I i n i-H a:Le d}

( l-) Th: tent ti ee presented here ere typicel of 0= ; plants with '.00 icgic a.cdificetier (eee Section 1.10. 2. II.K. 3.1 ? L

CO mai;._; 21;;:- i11-6 m. idad whou i.h. iiCG3 unipe 2000 analyci: i:-cu:,a.i t.;.ed in Ju ly 10 ^ ,.

Amendment 8

HCGS FSAR 8/83 OUESTION 440.0 (SECTION 6.3.3.7.3)

Provide the date when the plant-specific LOCA analysis will be submitted in an ammendent.

RESPONSE

The plant-specific.LOCA analysis 111 bc p c.ided i.- Jcly i^ 0 b"~

h a.5

. been com plea e cl . S e c +io n 63 has been' t' e v 65 cc) to p rovid c the. resalts o$ Phe M c6 S - speci fi c Q.n o t y 3 is -

3cR i f C rn C-1Q 440.0-1 Amendment 1

l I

HCGS FSAR 10/84 l )

l OUESTION 440.27 (SECTION 6.3)  !

I The references provided for the ECCS analysis must include '

references for the latest model changes and corrections used in the HCGS analysis.

i

RESPONSE

h o.s be e.n c.ornpleded .

The HCGS-specific ECCS analysis t il be precided ir. Bly !?SS andeY W utilizedthe LOCA evaluation models approved by the NRC in Reference 1 and described in Reference 2. i l

REFERENCES l

1. Letter to G. G. Sherwood (General Electric) from l R. L. Tedesco (NRC), " Acceptance for Referencing of Topica]  ;

Reports NEDE-20566P, NEDO-20566-1 Revision 1, and l NEDE-20566-4 Amendment 4," February 4, 1981. '

2. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K,"

NEDE-20566P, November 1975.

i 440.27-1 Amendment 8 l C -' Y scR iTC M

HCGS FSAR 12/83 OUESTION 440.28 (SECTION 6.3)

Justify ~ selection of a Lead plant for the LOCA break spectrum analysis. HCGS is committed to submit a plant specific LOCA analysis. We require a schedule for submittal of the plant specific LOCA analysis.

RESPONSE

The lead-plant LOCA analysis is an appropriate and representative break-spectrum analysis for the HCGS because the LOCA characteristics of BWR plants with similar ECCS configuration have been shown to be quite similar. The lead-plant analysis serves to identify the limiting failures and breaks and to

describe the general LOCA characteristics of these plants. Lead-plant sensitivity studies have demonstrated that the location of the limiting break is insensitive to slight variations in ECCS configuration and to changes in power level or fuel type. HCGS-

{ specific analyses will be provided at the limiting locations to define the specific HCGS response for the limiting cases. This is 'he c basis of the lead-plant concept.

l The  :: ult: 05MbeHCGS-specificECCSanalysis{aillbesebsitted

.bf. ._ $. N. 5. 2 .- .bc

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I 440.28-1 Amendment 3

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. - w SER Confirmatory Item No. 25 (SER Section 7.6.2)

"The staf f review of the elementary diagrams does not indicate that the EOC RPT transfers the pumps to low-frequency M/G sets af ter tripping their main power supplies. At previously review-ed BWRs (e.g. , Susquehanna (NUREG-0776) and River Bend (NUREG-0989), this transfer takes place af ter the RPT and the pumps run at approximately one-quarter their nonnal speed.

"There is not sufficient information for the staf f to c omple te its review regarding the EOC RPT. The applicant is required to submit design details showing the transfer of the recirculation pump power supply to a lower frequency motor / generator set upon EOC RPT. This is a conf irma to ry item. "

Response

End-of-cycle recirculation pump trip (EOC RPT) p'rovides for the insertion of negative core reactivity to improve therral margins for certain pressurization transients. The ef fectiveness of the EOC RPT arises from the rapid decrease in core flow that causes an increase in core voids bnmediately following the trips of the pump breakers. The early part of the transient and the core void reactivity the EOC RPT produces are not dependent on whether the final recirculation flow is determined by natural circulation or by a small power input to the recirculation pumps from a low-f requency motor / generator se t. None of the GE BWR/4 plants has installed a BWR/5/6-type of low-f requency M/G se t. Such installa-tions serve no safety f unction in the BWR/5/6 plants, and their absence is in no way detrimental to the ef fectivness of the EOC RPT for the BWR/4 plants. The above SER statement, which infers the existence of a low-f reque ncy M/G set for Susquehanna, is incorrect.

A .x r

SER Confirmatory Item No. 37 (SER Section 15.9.3)

.A plant-specific analysis must be provided to justify the bypass timer setting.

The staff finds the conceptual design for ADS logic modification proposed by the applicant acceptable confirmatory on completion of-the above specified actions.

Response

Plant-specific analyses have been completed to support Hope Creek's modified. ADS logic design that includes a bypass of the high drywell pressure trip after a sustained low water level signal and the addition of an ADS manual inhibit switch.

The analyses considered possible design requirements for both a minimum bypass timer setting consistent with ATWS considerations and Hope Creek's RRCS logic design and for a maximum setting based on ECCS performance evaluations. The ATWS evaluation determined that there would be no Level 1 interaction expected for postulated events. The results of the-ECCS evaluations are provided in response to confirma-tory. Item No. 14. These analyses are used to establish the technical specifications for the ADS timer and the bypass t ime r .

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