ML20138R733
| ML20138R733 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/31/1985 |
| From: | Heitner K Office of Nuclear Reactor Regulation |
| To: | Butcher E Office of Nuclear Reactor Regulation |
| References | |
| TAC-59787, NUDOCS 8511180816 | |
| Download: ML20138R733 (49) | |
Text
{{#Wiki_filter:_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ October 31, 1985 Docket No. 50-267 MEMORAN[UM FOR: Edward J. Butcher, Acting Chief Operating Reactors Branch No. 3, DL l i FROM: Kenneth L. Heitner, Project Manager Operating Reactors Branch No. 3, DL l l
SUBJECT:
SUMMARY
OF MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS SCHEDULE EXTENSION TO EQUIPMENT QUALIFICATION RULE (10 CFR 50.49) FOR FORT ST. VRAIN The purpose of the meeting was for PSC to make a presentation to the staff on their extension request. It also afforded PSC an opportunity to discuss with the staff their equipment qualification program, and the new steam line rupture detection and isolation system (SLRDIS). Attendees at this meeting are listed in Enclosure 1. The material used by the licensee in their presentations is in Enclosure 2. During these discussions, certain issues were noted for further attention by the staff and the licensee. These included:
- The aging methodology being used to qualify the equipment. - The adequacy of quality assurance in PSC's EQ program. - The spectrum of helium releases from the reactor or other pressurized helium lines should be considered by the PSC EQ program. - The environn'ents that could be created by small steam leaks that would not imediately be isolated by the SLRDIS. These environments could affect EQ or operator access following an accident.
The staff noted that additional information would be requested concerning these issues that affected both the extension request and.the entire EQ program. These questions would also cover:
- SLRDIS and, - Models used for determination of the accident environment.
The staff stated it would send a request for additional information to PSC shortly on these issues. Kenneth L. Heitner, Project Manager Operating Reactors Branch No. 3, DL
Enclosures:
As stated cc w/ enclosures: See next page ORB #3:DL N KHeitner;ef 8511180816 851031 to /3)/85 ADOCK O 27 fDR
r l k e MEETING
SUMMARY
DISTRIBUTION Licensee: Public Service Company of Colorado . _
~
- Copies also sent to those people on service (cc) list for subject plant (s).
Docket FU e; NRC PDR L PDR ORB #3 Rdg KHeitner EButcher BGrimes 0 ELD EJordan, IE ACRS-10 PMorriette NRC Meeting
Participants:
Glainas TKing RLaGrange PShemanski AMasciantonio JWermeil NWagner
*:Pt
. ATTENDEES AT OCTOBER 29, 1985 MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO ON EQUIPMENT QUALIFICATION AT FORT ST. VRAIN Name Organization ,
t Ken Heitner USNRC " E. Butcher USNRC G. Lainas USNRC T. King USNRC R. LaGrange USNRC P. Shemanski USNRC A. Masciantonio USNRC J. Wermiel USNRC N. Wagner USNRC S. Ball ORNL M. Harrington ORNL
- 0. Lee PSC L. Brey PSC M. Holmes PSC D. Warembourg PSC
. Mr. O. R. Lee Public Service Company of Colorado Fort St. Vrain l I
~~~'
cc: C. K. Millen Albert J. Hazle, Director Senior Vice President Radiation Control Division Public Service Company 4210 East lith Avenue of Colorado Denver, Colorado 80220 P. O. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein,14/159A Public Service Company of Colorado GA Technologies, Inc. P. O. Box 368 P. O. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0. Box 640 Platteville, Colorado 80651 Kelley, Stansfield & 0'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000
. Arlington, Texas 76011 su Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 1
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. Enclosure 2 EQ MEETING AGENDA 1
. OCTOBER 29, 1985 MEETING PSC - NRR I 1.0 INTRODUCTORY REMARKS - NRR f" PSC 2.0 SEPTEMBER 24, 1985 REQUEST FOR SCHEDULAR EXTENSION
- OVERVIEW, EXCEPTIONAL CIRCUMSTANCES
- RELATIONSHIP OF TECHNICAL ISSUES
- PSC POSITION FOR OPERATION BEYOND NOVEMBER 30, 1985 -
e M - SCH'EDULAR EXTENSION . CLARIFICATION NRC QUESTION 30 QUESTIONS / DISCUSSION
. s 4.0 ENGINEERING EVALUATION
- TEMPERATURE PROFILES
- SLRDIS
- OPERATION BEYOND NOVEMBER 30, 1985-(BRIEF ENGINEERING POSITION)
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5.0
SUMMARY
/ CLOSING REMARKS I
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9 EXCEPTIONAL CIRCUMSTANCES ,, . . di FORT ST. VRAIN IS A HIGH TEMPERATURE GAS-COOLED REACTQR, THE ONLY ONE OF ITS KIND IN THE USA. THEMAJORPOINTISTHATOURUNIQUECHARACTERISTICSDIdNOT l PERMIT PSC TO BENEFIT FROM INFORMATION SHARING AND GUIDANCE AVAILABLE IN THE LIGHT WATER REACTOR INDUSTRY . THROUGH ESTABLISHED OWNER'S GROUPS AND NRC INTERFACES PSC RECOGNIZED FSV'S UNIQUE HARSH ENVIRONMENTAL CONDITIONS AND PROCEEDED 10 DEVELOP AN EQ PROGRAM AND ASSOCIATED EQUIPMENT TEST PROGRAM TO QUALIFY EQUIPMENT TO THESE UNIQUE CONDITIONS ., E - IN - THIS RESPECT, FSV'S UNIQUENESS BY ITSELF DID NOT d_ DIRECTLY CAUSE EQ PROGRAM DELAYS, BUT INDIRECTLY THE
, UNIQUENESS WHICH TENDED TO ISOLATE FSV FROM INFORMATION AND GUIDANCE LED TO THE OVERALL EQ PROGRAM SHORTCOMINGS THAT WERE RECENTLY IDENTIFIED o
S
EXCEPTIONAL CIRCUMSTANCES GD THE EQ RULE (IOCFR50.49) DOES NOT RECOGNIZE THE UNIQUE CHARACTERISTICS OF AN HTGR. THERE WAS ESSENTIALLY NO HTGR, EQ , GUIDANCE AVAILABLE WITHIN THE NUCLEAR INDUSTRY g PSC WAS CHARGED BY AN NRC ORDER TO DEVELOP AN EQ PROGRAM UTILIZING DOR GUIDELINES TO THE EXTENT APPLICABLE TO AN HTGR PSC ACCEPTED THIS CHARGE AND PROCEEDED TO DEVELOP AN EQ PROGRAM BASED UPON CERTAIN ASSUMPTIONS. THESE ASSUMPTIONS AS WELL AS VARIOUS PSC POSITIONS WERE SUBMITTED TO THE NRC. PSC RESPONDED IN A TIMELY FASHION TO THE VARIOUS , -- BULLETINS, ORDERS AND REGULATIONS - (J- - NRC STAFF - WAS HEAVILY INVOLVED IN LWR EQ PROGRAM - l DEVELOPMENT. THE FSV EQ PROGRAM WAS DEVELOPED WITHOUT THE BENEFIT OF SIMILAR INVOLVEMENT OR FORMAL NRC REVIEWS J b i
, , , - - -- - . , , , . , -- , .,. , - - - - - . . ,-a- -n
O UNIQUE CHARACTERISTICS
- - NO CONTAINMENT
- NO RADIATION EFFECTS
~
- EMERGENCY RESPONSE SYSTEMS ARE IN OPERATION FOR NOR!blL PLAN -
OPERATION , t
- REACTOR CONCEPT UTILIZES A MULTITUDE OF EQUIPMENT TO EFFECT SAFE SHUTDOWN WITH SEVERAL SHUTDOWN METHODS 90 m on des ~ bo
- REACTOR CONCEPT ALLOWS TIME FOR OPERATOR ACTION
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- NO DESIGN BASIS EVENT' RESULTS IN OFFSITE RADIOLOGICAL cwli CONSEQUENCES THAT APPROA' i 10 CFR 100 GUIDELINES
- REACTOR CONCEPT ALLC'J FOR PERMANENT LOSS OF FORCED
- CIRCULATION WITH OFFSI' E RADIOLOGICAL CONSEQUENCES WELL BELOW ~
10 ,CFR 100 GUIDELINES C- . .
- HIGH TEMPERATURE /HIGH PRESSURE STEAM CONDITIONS RESULT IN EXTREMELY HARSH ENVIRONMENTS e
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, e EXCEPTIONAL CIRCUMSTANCES S RECENT NRC STAFF GUIDANCE HAS BEEN HELPFUL, BUT THE FSV PROGRAM HAS NEVER BEEN GIVEN A COMPREHENSIVE REVIEW WITH RESULTING ,SER OR TER
- PSC BELIEVES THAT MEANINGFUL AND CONSTRUCTIVE INTERACTIONS BETWEEN PSC AND THE NRC IN TERMS OF SER'S OR TER'S COULD HAVE RESULTED IN TIMELY DEVELOPMENT OF AN FSV EQ PROGRAM IN ACCORDANCE WITH 10CFR50.49
- IT IS EVIDENT THAT ONCE CONSTRUCTIVE INTERCHANGE WAS ESTABLISHED IN EARLY 1985. PSC HAS TAKEN AN AGGRESSIVE APPROACH TO RESOLVE ALL TECHNICAL ISSUES AND IDENTIFIED
_ CONCERNS -
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EXCEPTIONAL CIRCUMSTANCES ID BASIC ASSUMPTION AND LICENSING BASIS OF FSV EQ PROGRAM HAS BEEN UTILIZATION OF A FOUR MINUTE OPERATOR RESPONSE TO ISOLATE AN HELB. CURRENT NRC POLICY SETS FORTH A MINIMUM OF TEN MINpTES F FOR OPERATOR RESPONSE FSV HIGH TEMPERATURE / PRESSURE STEAM CONDITIONS ARE SUCH THAT AN EQ PROGRAM CANNOT BE DEVELOPED FOR A TEN MINUTE OPERATOR RESPONSE , RESOLUTION OF THIS ISSUE (IN!TALLATION OF A STEAM LINE RUPTURE DETECTION / ISOLATION SYSTEM) HAS A BROAD OVERALL IMPACT ON THE EQ PROGRAM I_ - ALTHOUGH NOT SPECIFICALLY STATED IN THE SEPTEMBER 24, 1985
- dD' SCHEDULAR EXTENSION REQUEST THE FOUR MINUTE VERSUS TEN .
MINUTE OPERATOR RESPONSE CANNOT BE DIVORCED FROM THE OTHER TECHNICAL ISSUES RAISED BY THE NRC IN JANUARY 1985 (I.E. AGING,OPERABILITYTIME,EQFILES,ETC.)
- THE GENERAL RELATIONSHIPS AND INTERDEPENDENCY OF OPERATOR RESPONSE TIME, AGING AND OPERABILITY TIME ARE REPRESENTED ON THE FOLLOWING TWO TABLES 6
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P , 'l I ,e
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e I I f Eg ISSUE RESOLUTION ACTIVITIES / EwtLUATIONS E0 lerACT l I EvnLUATE/DETEneMNE BASIS ORIGNAL EQ PROGRAM DID
' FOR ENVIROpedENTAL CONDITIONS NOT CONSIDER AGING S I AND AGING EXPERIENCED TO DdtE OPERASILITY TIME. ENTIRE
' PROGRAM IS IMRACTED FIELD IIIALKDOWN & VOIEFICATION TO ASCERTAIN COMPLETE MODEL PROGR448 IMPACTED BY SUtDIS 8
Nues8ERS S OTHER INFO BECAUSE OF CHANGES IN
' ENvlRONGENT CONDITIOe:S '?
DETERMiest / IDENTIFY AGE e SENSITIVE MATERIALS NEW EQUIPMENT INTRODUED
- BY SLROIS
? CONDUCT STUDIES AND
! EVALUATIONS TO INCOIWOfUCT AGGIS/Opraansa ITY TadE AGNIG Ale 0 OPERASILITY TesE DETEllediteE 90EW HARSH
} IseTO THE OVERALL E3 AGE SENSITIVE MATERIALS
, 1 NOT III EQ PROWIAM ENviflO88 MENT CONDITIONS
# FOR AGING CALLdLATIONS OR EQUIPMENT 880ST BE
- REPLACED 1
l DETUteAIIeE ACCESS 180LITY FOR EQUIP REPAIR FOR OPERA 8ellTY AD0fTIOral PREVIOUSLY UNNNOt% COceDITIOseS I CONSIDERATIONS 1 ' IDENTIFIED IN FIELD WALMDOWN ItEPACTED PROGRAM (le TAPED SPLICES)
- IDENTIFY AT RISK EQUIPteENT Ape 0 EVALUATE /DETERetWIE I CORRECTIVE ACTION EQ FILES FOR ENTIRE l PROGRAM 18UST BE j
' + DEVELOP OVER ALL Pts PROGRAM _:. FOR AGE SENSITIVE EQUIPedENT , i SYSTEMS /COe8PONENT REVIEW TO DETERe8IIC ELECTRICAL STATE ' ASSOCIATED feON-SAFETY MED t$~ ITEMS & CYCLIC OPER FOR J AGING AND OPERA 81LITY ,bW @ (Othe @ bC
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q, i . I i ACTMTIES / EVALUATIONS EQ IMPecT EQ ISSUE RESOLUTION GENERATE DIE 11r TEarERATURE REQUIRES EQUIR RE-EVALUATION PROFILES & ASSOC, CONDITIONS FOR NEW ENVIRONMENTAL CONDITIONS / OPERATOR ACTION SYSTEM DESIGN
. REQUIRES EQUIR NOT
-CROR NePeCT PREVIOUSLY OUALIFIED SOteE OF i
- RELIA 8ILITY WHICH IS NON-SAFETY RELATED I -OTHER DESIGN BASIS EVDITS
, REQUIRES 98000FICATION OF 4 MINUTE VS to teessuTE INSTALL STEAM LINE I ' EQUIPedENT COpeFIGURATIO88 OPERATOR RESPONSE RUPTURE DETECTION . I MATERIAL PROCURE 0ENT I Aeso CIRCulTRY J~ ISOLATION SYSTEtt
- (SLROISI FIELD IIghlKDOWN/ VERIFICATION MS MWEE M i
FOR NEW EQUseteENT PPS 8 AW CMTm EOur LOCATION
! IsODEL NUnd8ERS I FIELD CONFIGURATIODI REQueRES 9800lFICATIO86 AS A POWER SOURCE RESULT OF OTHER ACCIDENTS 1
OR CONDITIO8sS i ACCIDENT A88ALYSES ANO SAFETY EVALUATION AGeseG PROGRAM les%CTED I BY ENVIROosteENTAL CONomONS OPERATiteG / EndERGEleCY PROCEDURES OVERALL EQ PROGRAAA DOCuteENTATI(NI FOR ENTIRE PROGRene' i l tsOctFIC ATIOft / INSTALLATIONl ADDITIOesAL,PREVIOUSLY Use880WN. CO800lTIONS IDENTIFIED SY FELD WALKDOwhS Afe0 OTHER STUDIES a f 1
I, TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM
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EQ RELATED TASKS SE EO E N 55 E' U f 8 N ENGINEERING EVALUATIONS / ANALYSIS EE-EQ-001 I X I FSV EQ Functional Descriptions - EE-EQ-002 I X I
- FSV Component Operating Cycles ~_ EE-EQ-003 X C I X =
, g=- Interactioh and Effects of Non- . Qualified Comp. in Circuits with Qualified SSD Comp. EE-EQ-004 X C I X X Interaction and Effects of Non-Qualified Comp. in Circuits with Qualified P?3 Comp.
,' , EE-EQ-005 . X C I I Acceptability of Seismic Type D o1 - Valves in a Steam Break Environment EE-EQ-006 X C X Effect of. Fire Suppression System Actuation on Safe Shutdown Equipment EE-EQ-007 X C I X Interaction and Effects on Non- . Qualified Comp. in Indicator Light Circuits with Qualified SSD Comp.
EE-EQ-008 X C I X C C FSV Brake and Seal System Solenoid Valve Qualification EE-EQ-009 C I X C C Liner Cooling Method for SSD Cooling at FSV Following High Energy Line
- Break-EE-EQ-010 X X Potential, Failure of Components Subjected to' HELB Envtronment EE-EQ-011 X C I C C Systems Review Forced Lirculation '
Cooling
; TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM eE E a 5; e x 5 .
38 % ud 5 5 2" a %; o 1;
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SE T 55 E' E EQ F. ELATED TASKS EO E N UI ~gt A d ENGINEERING EVALUATIONS / STUDIES EE-EQ-013 I I I X I Accidents Resulting in Harsh Environment Areas at Fort St. Vrain EE-21-011 X C
- Submergence Concerns on BWST Level -
- Valves Provide HELB Profiles for Manual I I I X Termination of Feedwater and Condensate Breaks Evaluate Other Trip Methods for I I X I Feedwater Condensate and Extraction Steam HELBS
- ]
Provide HELB Profiles Utilizing I I I X I Variable Heat Transfer Coefficients Prepare Report on Blowdown Rates and X Building Air Temperature Response Evaluatio'n - Reactor Building Louvers C X and Exhaust Filters Revise CONTEMPT G to Provide Humidity I X I I
. Calculation Identify Control Functions Taken X X X X Credit for in the HELB Scenarios Prepare HELB Systems Review Document I I X I Including Single Failures Investigate Recalculation of MCA & I I I X DBA 2 Profiles with CONTEMPT G Taking Credit for Heat Sinks .
Revise CONTEMPT G to Provide Harsh I I I X i Environment Profiles for East 4A Wall, Turbine Deck and Refueling Floor
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TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM e e E E
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EQ RELATED TASKS ENGINEERING EVALUATIONS / STUDIES Investigate Min./ Max. Human Tolerance I I X C for Personnel Entering Harsh Environments Evaluate Automation of all Manual C X
~~
Valves Associated with Liner Cooling ,
= -
E- Evaluate Automation'of all Valves X C s Associated with Forced Cooling FIELD WALKDOWNS/ REPAIRS Perform Walkdown of Safe Shutdown I X I I I Items Identify or Replace ASCO Solenoids I X I I without Nameplates I I I Review Field Walkdown Sheets X Prepare Report on Model # Discrepancies Review Work Completed /c.' in Progress I X I I Since February 1983 [ Install New Splices I X I I I Perform Walkdown of Cables in EQ I X I I I Program as Modified by Cable P.O. Review Repair Items Identified by Walkdown I X I I
', TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM i
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S I' = f{ E* E EQ RELATED TASKS EO E Eo 3I *ftE d DESIGN CHANGES CN Package - Hard Pipe Spool Pieces X C X and Provide Firewater Piping to HTFA, CN-2142 CN Package - Protective Equipment X I X C C Changes to Insure Proper Coordination - Between SR,and NSR Loads, CN-2119 4-_ Replacement of Brake and Seal Valve X I X C C Seats and Seals, CN-2178 Prepare System Description for Steam X Line Rupture Detection / Isolation 4 System (SLRDIS) Tagging of Components (Indicating X X Lights) not Previously Tagged CN-2184 SLRDIS Modifications for E-H Valves. I I I C X (PV-2229,2230: PV-2243,2244), CN-2180
~
PPS Modifications for SLRDIS, CN-2176 X
- - SLRDIS Modifications for Air Operated I C I I C X and Hydraulic Valves (18 Valves),
= CN-2181 m Cable Re-routing for Components X C X Affected by EQ Program Changes, i,
CN-2182 CN Package - Move TR 8201, TR 8202, I C X I C C and TR 8203 to Mild Environment,
- CN-2151 CN Package - Replace GE Relays with I C X ! C C i Qualified Devices CN-2095 CN Package - Replace United Electric I C X C C Temperature Switches with Qualified Devices. CN-2160 CM Package - Replace Foxboro TE's with I C X C C
- Qualified Devices, CN-2144 '
, TASKS BEING PERFORMED TO SUPPORT THE FSV EQ PROGRAM m c E5 5 E e5 3 D ; t to ; 2; 8 3.
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5 EQ RELATED TASKS DESIGN CHANGES (CONTINUED) CN Package - Replace ASCO's for I C X C C FV-21297 and FV-21298 with Qualified Devices CN-2152 E
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. DATA BASE UPDATING ,
f_. EQ Data Base Update for Forced X C C X
, Circulation Cooling, Ten Change Notices Overall EQ Data Base Update, CN-2172 X I X I Preparation of EQ Books I C X I C C ,
Prepare Detailed List of Equipment X X X Required for Isolation of High Energy Line Break and Equipment Required for Depressurization and Liner Cooling Identify EQ Cable Numbers I X I I
- Provide Necessary EQ Documentation C C X C C for Liner Cooling Comp.
MISCELLANEOUS Transmit Voltage Ranges for EQ X Electrical Equipment to S&L Environmentally Test Fuse Blocks X I X C C for CN-2119 l LEGEND X - Primary initiating areas / functions I - Areas / functions impacted C - Areas / functions needed for completion / input 1
JUSTIFICATION FOR CONTINUED OPERATION O INSTALL SLRDIS O UTILIZE QUALIFIED REDUNDANT LINER COOLING PATHS ALONG WITH ASSOCIATED SYSTEMS TO PROVIDE REACTOR COOLING 0 PSC CONTENDS THAT INSTALLATION OF SLRDIS ALONG WITH QUA$1FICATION OF THE LINER COOLING FLOW PATHS AND ASSOCIATED SYSTEMS .WILL ESTABLISH NECESSARY CONDITIONS TO PROTECT THE HEALTH AND SAFETY
. OF THE PUBLIC AND MEET THE INTENT OF 10 CFR 50.119 FOR INTERIM 2 OPERATION (NOVEMBER 30, 1985 THRU MARCH 31, 1986) IN THAT:
i
- THE INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY CAN BE MAINTAINED ! - THE REACTOR CAN BE SHUTDOWN AND MAINTAINED IN A SHUTDOWN CONDITION .
h - THE CONSEQUENCES OF REACTOR COOLING UTILIZING THE PCRV LINER COOLING PATHS HAVE BEEN ANALYZED, AND IT HAS BEEN SHOWN THAT OFFSITE EXPOSURES REMAIN WELL BELOW 10 CFR 100 GUIDELINES O THE TECHNICAL ASPECTS AND SUPPORT FOR THE ABOVE POSITION WILL BE ADDRESSED BY ENGINEERING EVALUATIONS AND ASSOCIATED SAFETY EVALUATIONS FOR SLRDIS AND LINER COOLING TO BE SUBMITTED FOR NRC REVIEW IN EARLY NOVEMBER, 1985 - i
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1 EQ SCHEDULAR EXTENSION
> - PSC IS PROCEEDING WITH AN AGGRESSIVE EFFORT TO QUALIFY ALL ELECTRICAL EQUIPMENT NECESSARY FOR SAFE SHUTDOWN UTILIZING BOTH THE LINER COOLING METHOD AS WELL AS FORCED CIRCULATION
- PSC IS CONCENTRATING INITIAL EFFORTS ON EQUIPMENT NEEDED FOR SLRDIS AND LINER COOLING UTILIZING FIRE WATER, CURRENTLY THIS SET OF EQUIPMENT INCLUDES 51 EQUIPMENT ITEMS COMPRISING 438 COMPONENTS l
- PARALLEL EFFORTS ARE IN PROGRESS FOR QUALIFICATION OF ELECTRICAL EQUIPMENT NEEDED FOR FORCED CIRCULATION COOLING.
THIS EQUIPMENT SET PRESENTLY INCLUDES 156 ITEMS REPRESENTING
! 1,246 COMPONENTS
- PSC IS CONFIDENT THAT THE MAJORITY OF EQUIPMENT IN THE OVERALL
- EQ PROGRAM WILL BE QUALIFIED IN THE NOVEMBER 30, 1985 OR PRIOR -
- 70. REACTOR RESTART TIME FRAME INCLUDING REPLACEMENT OF AGE *
, E_- SENSITIVE MATERIALS / COMPONENTS, REPAIR OF FIELD INSTALLATION i ,
DEFICIENCIES AND REPLACEMENT OF TAPED SPLICES
- IT SHOULD BE EVIDENT, HOWEVER, THAT PSC IS INVOLVED IN A MULTI-FACETED EFFORT WITH MANY INTERDEPENDENT ACTIVITIES PROCEEDING IN PARALLEL. SIGNIFICANT PROBLEMS THAT COULD DEVELOP IN ANY ONE AREA COULD AFFECT MANY OTHER AREAS AND ; COULD HAVE NEGATIVE IMPACT ON THE OVERALL EQ PROGRAM SCHEDULES.
3 PSC HAS EVALUATED THE CURRENT STATUS OF EVALUATIONS / STUDIES AND OVERALL EQ PROGRAM WORK ACTIVITIES. PSC BELIEVES THAT NECESSARY ACTIONS REQUIRED TO ESTABLISH A DEFENSE IN-DEPTH EQ PROGRAM CAN BE ACCOMPLISHED WITHIN THE MARCH 31, 1986 SCHEDULAR EXTENSIONS REQUEST. I l . a i m - ,
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- PERSONNEL CURRENTLY WORKING ON EQ ,
PSC AND ON-SITE CONSULTANTS FOR ALL PSC DIVISIONS (PRODUCTION, ENGINEERING, LICENSING, QA), EQ IS TOP PRIORITY AND USES MAJORITY OF WORK EFFORT. PSC HAS CONTRACTED NUMEROUS CONSULTANTS TO WORK ON SITE TO SUPPLEMENT PSC STAFF WORKING ON EQ
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0FF-SITE CONSULTANTS ,
-T -
! dC- DI BENEDETTO ASSOCIATES GA TECHNOLOGIES INC GENERAL ELECTRIC IMPELL PROTO-POWER CORPORATION SARGENT & LUNDY 1 STONE & WEBSTER TENERA , WESTINGHOUSE
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WYLE TOTAL THE ABOVE WORK FORCE REPRESENTS THE EQUIVALENT OF OVER 400 PEOPLE WORKING FULL TIME ON EQ 1
,i i l N,RC QUESTION #1 1 O WHAT IS THE CAPABILITY OF THE REACTOR BUILDING CONFINEMENT AND LOUVERS TO WITHSTAND A STEAM LINE BREAK AND STILL FUNCTION EFFECTIVELY TO REDUCE THE RADIOLOGICAL CONSEQUENCES OF THE. DESIGN i BASIS EVENT? e IN TERMS OF OFFSITE RADIOLOGICAL CONSEQUENCES. TH$ VARIOUS DESIGN BASIS EVENTS AS CONTAINED IN THE FSV FSAR DOFNOT TAKE CREDIT FOR TFE RE/CTOR BUILDING LOUVER SYSTEM
- THE REACTOR BUILDING LOUVER SYSTEM AS IT CURRENTLY EXISTS IS I NOT ENVIRONMENTALLY QUALIFIED I-e EVALUATIONS ARE UNDERWAY FOR THE FOLLOWING AVENUES OF APPROACH:
- DETERMINE THE END EFFECT OF SLRDIS AND THE NEED TO QUALIFY --$y(
THE LOUVER SYSTEM g=-
- h e,nh}-
=
- QUALIFY THE REACTOR BUILDING LOUVER SYSTEM WITHIN THE AUSPICES OF FSV'S EQ PROGRAM
- PERFORM EVALUATIONS AND ANALYSES TO SUPPORT NOT UTILIZING OR DEPENDING ON LOUVER SYSTEM ACTUATION 1
NRC QUESTION #2 I 0 WHAT IS THE EFFECT OF MOISTURE RELEASED BY 1HE BREAK ON THE EFFECTIVENEJS OF THE REACTOR BUILDING EXHAUST FILTERS?
~
- PRELIMINARY INVESTIGATIONS INDICATE THAT THE PERFORMANCE OF THE EXHAUST FILTERS WILL NOT BE SIGNIFICANTLY REDUCED BY THE STEAM LEAK CONDITIONS IN TERMS OF EFFICIENCY FOR REMOVAL OF ELEMENTAL IODINE FILTER EFFICIENCY COULD BE REDUCED FOR ORGANIC HALIDES
- =
- PSC IS CURRENTLY EVALUATING THESE AREAS IN TERMS OF RADIOLOGICAL CONSEQUENCES. PRELIMINARY BUILDING TEMPERATURE PROFILES AND RELATIVE HUMIDITY PROFILES HAVE BEEN RECENTLY COMPLETED AND ARE UNDERGOING FINAL REVIEW. THIS INFORMATION
; WILL BE FACTORED INTO THE EXHAUST FILTER EVALUATIONS TO 1 DETERMINE IMPACT
- EVALUATIONS AND ADDITIONAL INFORMATION SHOULD BE AVAILABLE EARLY IN NOVEMBER Locht oh 3 l l
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NRC QUESTION #3 GD WHAT ARE THE SAFETY BENEFITS FROM OPERATING THE PLANT AT PARTIAL POWER DURING PERIOD OF NOVEMBER 30, 1985 TO MARCH 31, 1986~ . L
- FROM THE VIEWPOINT OF RESULTING EQ CONDITIONS FROM AN HELB, THERE IS ESSENTIALLY NO BENEFIT OPERATING AT PARTIAL POWER LEVELS PSC HAS NOT PERFORMED ANY DETAILED ANALYSIS TO QUANTIFY
- REDUCED RADIOLOGICAL CONSEQUENCES DUE TO A LOWER FISSION PRODUCT INVENTORY OF LOWER FUEL TEMPERATURES. SOME RADIOLOGICAL BENEFIT COULD OBVIOUSLY BE EXPERIENCED
-- - THE OAK RIDGE FSV SCENARIO OF OCTOBER 14, 1985 FOR VARIATIONS -
0F DBA.-1 INDICATES FUEL FAILURE WILL OCCUR FROM POWER LEVELS (." 35% AND AB0VE. BUT THAT THE PERCENTAGE OF FUEL FAILURE AT 5 - DAYS IS SMALL AND THEREFORE FISSION PRODUCT RELEASE IS SMALLER 4 S I e4 I i . s
_ c . . . .... o OAK RIDGE NATIONAL LABORATORY *' *** *o= v oaa acos. re.ases mn openarWD ev annarse hammeTin megy sysfaa sec. October 14, 1985 b*b kk kC \O $ s x.,sm Mr. T. L. King emt. Office of Nuclear Rasetor Regulation ' h,y'. 4. . s. Mail Stop P846 g,su,; cap. U. S. Nuclear Regulatory Commission gyo;j.. J. Washington, D.C. 20555 pgm I [-
Dear Tom:
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ORICA Code Calculations of Fort St. Vrain DBA-1 Scenario Variations . mesiap. a. Per your request, I have enclosed results of analyses of Fort St. Vrain E=h. 5. s. (FSV DBA-1 scenarios for less-than-full power operation (see Table 1). EEE'ai/* The stvere accident version of the ORECA-FSV code was used for these Eh i calculations. The main object was to determine the maximum equilibrium E E .'is.nostE power level FSV could run at such that if a permanent loss of forced Ei. circulation (LOFC) event (DBA-1) did occur, the maximum fuel temperature would be such that essentially no fuel damage would occur, i.e. T-core (max) <2900'T or 1600'C. The study showed that this power level is %35% [or 295 MW(c)]. The reference case assumptions for the parameter study were:
- 1) Primary system scram, depressurization, and 14FC occur at time = 0.
Only one (of the 2) liner cooling systems (LCSs) is opera-
~
2) o tional.
- 3) The initial core average temperatures assumed are (conser-vatively) derived from FSV normal operating data (see Fig. 1).
- 4) A time-at-temperature fuel failure model is used to estimate fuel failure fractions *.
Sensitivity studies were also done. It was seen from them that the maximum core temperatures were only very slightly less both for the case where both LCS trains are operational and for the case where the system is depressurized according to the emergency operating procedures (EOP-G), i.e., begin a 7-hr depressurization 2 hours af ter the 14FC. D. T. Goodin, " Accident Conditf.an Performance of Fuels for HTGRs", J. Amer. Ceramic Soc., 65,, 5, May 1982, pp. 238-242.
Mr. T. L. King 2 October 14, 1985 Several cases were also run'in which it was assumed that no depressurization occurred (and the pressure was assumed to remain constant). Becau_se of the enhanced convection heat transfer to the LCS, the maximum core ~
- ' temperatures are considerably less for corresponding powere for.the depressurized cases. While such a mode of operation is contrary to the 1 current tech specs and not reconsended, the analyses indicate thai for l operation up to 40% power, a DBA-1 scenario in which depressurization l was not achieved would not lead to a situation in which a subsequent l 1
rapid depressurization would result in the spread of significant fission products from failed fuel. l 1 l The DBA-1 scenario accounts for overheating and eventual failure of the l I PCRV's carbon steel cover plates, especially in the upper plenum region. ; Noting that the ORECA models used are " approximate", the results indicated l l that the maximum cover plate temperatures peaked at 1200'F and 1500*F. i respectively, for the 20% and 30% power depressurized cases. Hence ] l beyond 30% power, significant cover plate failure would be pred1cted.
, 4" , ,
Two other comuments on the results presented: i
- 1) The Goodin fuel failure model used was reviewed by ORNL and
' judged to be excellent for predicting relatively large failure
' fractions, but it was not intended as a means for predicting onset-of-failure situations such as we have here. Bence, j while we believe that these results are good for ballpark j estimations, they are not accurate. 4 t ] 2) The 1600'C fuel limit is a convenient yardstick, but the , Goodin model is better, and shows both that some failure occurs at lower temperatures and that not necessarily much more occurs at the higher temperatures reached, for example, in the 402 power depressurised case (Table 1). Note also that the fuel failure fraction is less than the fraction of fission i j' product release to the primary system. The model does not
, calculate diffusion through the graphite element, etc.
i Please let us know if you have questions or comments. Yours truly, '
- 5. J. Ball, Manager BTCR Safety Studies for NRC i Enclosure l ces J. C. Cleveland T. S. Kress l
R. 3. Foulds - RES A. F. Malinauskas ! R. M. Barrington D. L. Moses l K. L. Neitner - NRR J. E. Wilson N. B. Holmes - PSC l R. F. Wichner ; R. E. Ireland - R4 ! 1
' t i
1
- _ _ _ . - _ _ , - . , . _ ~ . , _
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TABLE 1. FSV DBA-1 SCENARIO VARIATIONS INITIAL T I DEPRESSURIZED POWER CORE (MAX) PEAK TIME FUEL FAILURE
- CASES LEVEL 'F ['C] DAYS 9 t=5 DAYS 20% 2180 1193 3.5 0 30% 2681 1472 3.5 0.04%
35% 2902 1594 3.4 0.28% 40% 3110 1710 3.1 0.75% PRESSURIZED _ CASES -
- 40% 2520 1382 1.25 0 *
~
(P= CONSTANT) 50% 2795 1535 1.25 0.06%
~
*G00 DIN (CAT) MODEL USED. INCLUDES TIME 9 TDtP. MODEL, NOT ACCURATE 9 LOW FF'S.
' NOTE FAILURE % AT 40% POWER VERY SMALL. FP RELEASE IS SMALLER.
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NRC QUESTION #5 O WHAT ACTIONS WOULD BE REQUIRED TO RETURN FSV TO POWER OPERATION FROM A PERMANENT LOSS OF FORCED CIRCULATION COOLING FROM PARTIAL POWER CONDITIONS?
- PSC HAS NOT PERFORMED ANY DETAILED ANALYSES CONCERNING FUEL TEMPERATURES AND ULTI!! ATE PCRV INTERNAL TEMPERATURES FOR A PERMANENT LOFC FROM PARTIAL POWER LEVELS g
- BASED ON ENGINEERING JUDGMENT, WITH NO OTHER CLAIMS WHATSOEVER, OPERATION AT ANY SIGNIFICANT POWER LEVEL THAT WOULD REPRESEAT ECONOMIC BENEFIT WOULD IN ALL LIKELIHOOD RESULT IN PERMANENT PCRV INTERNAL COMPONENT DAMAGE. THE ORNL STUDY OF OCTOBER 14, 1985 INDICATES SUPPORT OF THIS CONCLUSION
[_ . . . s s
s STEAM LINE RUPTURE DETECTION AND ISOLATION SYSTEM (SLRDIS) PURPOSE Provide Continuous Monitoring of Area Temperatures in both Reactor & Turbine Buildings
- Minimize SulldingEnvironmentalConditionsfollowingkteamLine Rupture to:
I Protect Functional Integrity of Safe Shutdown Equipment Enhance Re-entry into Plant Areas for Recovery SYSTEM SCOPE Temperature Sensors (Thermistor Cabling) - C - d'J "
- Microprocessor Logic Cabinet Interface with PPS for:
- Circulator Trip (1) Four Circulator Trip (2) Two Loop Trouble (3) Reactor Scram Valve Actuation b
l i.s 1 s
DESIGN FEATURES / BASES
- Capability to Detect: - -
~
High Temperature Conditions - (1) Pre-alarm 9 150-200 degrees F (2) Trip 9 250 degrees F (3) Rate of Rise Alann 9100 degrees F/ min. Trouble -
~
Ability to Withstand Single Active Failure , e p -
- 2/4 Logic in Sensing Circuits s
Redundant Microprocessors Redundant PPS Logic Redundancy 0 Valve Actuations
- Capability to Function Without Offsite Power Necessary to Operate Without Operator Action for First Ten (10)
Minutes Operate in parallel with, but independent from existing Steam Pipe Rupture Detection System Leave Existing SPRDS Intact Keep Existing Tech. Specs. -
- kAca%cw l
1
SYSTEMFUNCTIbNS
- Monitor R8 & TB Area Temperatures .
- RB ,.
- TB
- Monitor Each Zone Redundantly
- Four (4) Sensors per Zone Redundant Microprocessors
- 2/4 Logic Trip Signal to PPS
~
- Isolate Secondary Coolant & Power Conversion System (via PPS)
- f=-
~
Isolate FW (Yalves 2, 3, & 4) to Eliminate Steam Production
- Isolate MS (Valves 7, 8, & 9) to Eliminate Steam Flow
- Isolate RH (Valve 15, 16, & 17) to Eliminate Steam Flow
- Trip & Isolate Circulators (Valves 13 & 14, also Water Turbin.e Valves not shown) to Eliminate Helium Flow thru SG Isolate Aux. Boiler (Valves 20,21,22,31) to Eliminate Steam Flow Initiate Reactor Scram (via PPS) ,
SYSTEM RESET & RECOVERY
- Consistent with Existing Reset Procedures, to Maximum Extent a Possible
- Microprocessor Reset
- PPS Reset
- Circulator Trip Valve Actuation (1) Reset 1 (2) Reset 2 l
FIGURE C: SPRDS, SLRDIS, AND PPS RELATIONSHIP ( A LOGIC SHOWN, B LOGIC SIMILAR) INPUTS FROM INPUTS FROM SPRDS SLRDIS INSTRUMENTATION . INSTRUMENTATION y ^ 000 ' 0000 ( I Nk tW I Thy l 1r 1r 1r 1r1rif.1P 4" SPRDS STEAM LINE )
-LOGIC CIRCULATOR RUPTURE -
TRIP ((ONE PER CIRCULATOR) DETECTION LOGIC ', RACK l (I-93543) { PPS ', VALVE i ACTUATION = l LOGIC l a i
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'91 SORS TEMPERATURE l SENSING CONTROL ROOM jy.
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O TABLE 1
- TAG SLRDIS ACTUATION RESET NO. LOOP DESCRIPTION LOGIC METHOD (1) ACTION SV-2105 1 CIRC 1A SPEED CONT A&B CT , (2) i SV-2106 2 CIRC IC SPEED CONT A&B CT r" l SV-2109 1 CIRC 1A WATER TURB CONT A&B CT SV-2110 2 CIRC IC WATER TURB CONT A&B CT l
SV-2111 1 CIRC IB SPEED CONT A&B CT SV-2112 2 CIRC ID SPEED CONT A&B CT SV-2115 1 CIRC IB WATER TURB CONT A&B CT SV-2116 2 CIRC ID WATER TURB CONT A&B CT HV-2109-1 1 CIRC 1A WATER TURB SUP A&B CT (2) HV-2110-1 2 CIRC IC WATER TURB SUP A&B CT
- HV-2115-1 1 CIRC IB WATER TURB SUP A&B CT HV-2116-1 2 , CIRC ID WATER TURB SUP A&B CT C_-
HV-2109-2 1 CIRC 1A WATER TURB DISCH A&B CT (2) HV-2110-2 2 CIRC IC WATER TURB DISCH A&B CT HV-2115-2 1 CIRC IB WATER TURB DISCH A&B CT HV-2116-2 2 CIRC ID WATER TURB DISCH A&B CT HV-2201 1 FW INLET B XCR ' (2) HV-2202 2 FW INLET A XCR I FV-2205 1 FW CONTROL A XCR (2) 4 FV-2206 2 FW CONTROL B XCR HV-2203 1 EMER FW INLET B XCR (2) HV-2204 2 EMER FW INLET A XCR HV-2223 1 SHT STM STOP CHECK A&B CT (2) HV-2224 2 SHT STM STOP CHECK A8B CT
~
PV-2229. 1 SHT STM BYPASS A&B XCR (3) PV-2230 2 SHT STM BYPASS A&B XCR i HV-2292 1 SHT STM STARTUP BYPASS A&B XCR (3) HV-2293 2 SHT STM STARTUP BYPASS A&B XCR l HV-2241 1 RHT STM BYPASS A&B XCR (3) HV-2242 2 RHT STM BYPASS A&B XCR PV-2243 1 RHT STM BYP PRESS RATIO CONT A XCR (3) PV-2244 2 RHT STM BYP PRESS RATIO CONT B XCR HV-2249 1 CIRC 1A TURB TRIP A&B CT (2) HV-2;f50 2 CIRC IC TURB TRIP A&B CT HV-2251 1 CIRC IB TURB TRIP A&B CT !
" l HV-2il52 2 CIRC ID TURB TRIP A&B CT HV-2253 1 RHT STOP-CHECK A&B XCR (3) i HV-2254 2 RHT STOP-CHECK A&B XCR
{
. TABLE 1 continued TAG SLRDIS ACTUATION RESET NO. LOOP DESCRIPTION LOGIC METHOD (1) ACTION PCV-5201 -
AUX STM TO 150 PSIG HDR A&B XCR ~ (3) PCV-5213 " AUX STM TO CRH A&B XCR . PCV-5214-1 PCV-5214-2 CRH TO 150 PSIG HRD CRH to 150 PSIG HDR A&B XCR k(3) A&B XCR PCV-5214-3 - CRH TO 150 PSIG HDR A&B XCR PCV-5305 - 150 PSIG HDR TO DA A&B XCR (3) (1) XCR indicates valve is actuated thru an XCR in PPS. CT indicates valve is actuated thru Circulator Trip Logic portion of PPS.
~
5I- (2) Requires: - s a. Return of ambient to below setpoint.
.b . Reset of microprocessor 9 Control Panel
- c. Reset via exiting methods to recover from Circulator Trip (3) Requires:
- a. Return of ambient to below setpoint
- b. Reset of microprocessor 9 Control Panel
- c. Reset of XCR via HS-93375 (HS-93376) on MCB i O
b 4 i i i e k
--_ _____a
4 s TEMPERAr!!RE PROFILES Using Bulk Building Temperature Curves . Consistent with nuclear industry - 1 Containment of water reactors l F l i 135'F Maximum Temperature for Access / Manual Actions Survey EPRI Report NP2868 i Developed Scenarios Offset rupture b Single-active failure f-
- Existing plant p'rotection and control systems
~
s SLRDIS Safety-Related Equipment Non Safety-Related Equipment SLRDIS
~*
Detectors Actuation times Profiles Blowdown Rates - FLASH Building Temperatures - CONTEMPT-G ' Results I l
I
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t e-l ENGINEERING EVALUATION OF i-LINER COOLING METHOD WITH FIREWATER _ FOR SHUTDOWN COOLING OF FORT ST. VRAIN - e
- 4i FOLLOWING A HIGH ENERGY LINE BREAE a ~
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PURPOSE OF THE ENGINEERING EVALUATION Document acceptability of PCRV liner cooling using firewater as-interim 7 method of decay heat removal following a high energy line break (HELB). Identify specific systems and components to be utilized? including electrical components in the harsh environment which must function, and define manual actions required. W ,
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BASIS FOR LINER COOLING METHOD Liner cooling with firewater to be used as the cesign basis for decay heat removal (DHR) following a HELB during interie period until forced circulation systems / components are qualified. . 8 I Liner cooling with firewater is used to minimize reliance on electrical components in harsh environments. Liner cooling with loss of forced circulation (LOFC) is addressed in FSAR as Design Basis Accident (DBA) No. 1. Liner cooling for DHR based on worst case scenario: . e p --Double ended., offset rupture of steam line . s --No credit taken for non-qualified electrical components located in harsh environment.
--Coincident loss of offsite electrical power
--Due to nonqualified components, systems for forced circulation are assumed to be lost.
--Liner cooling ~ is last option for core cooling. It can be expected that many forced circulation components, although not qualified, will survive the HELB.
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OVERALL PROCEDURE Depressurization of PCRV - .
-- To limit heat transfer from core to PCRV internals. ~,
-- Via Helium Purification System (HTFA, HPC, LTA) to Reactor Building exhaust system.
-- Must be initiated within two hours after LOFC to maintain acceptable heat load on Purification System components.
Actuation of Reserve Shutdown System
-- -- To assure adequate shutdown margin throughout the event. -
d7' -- Must be initiated within 5 hours. s PCRV Liner Cooling
-- Decay Heat Removal
-- DBA No. I assumes Reactor Plant Cooling Water System is available.
Use of firewater is addressed as option.
-- Must be initiated within thirty (30) hours after LOFC.
Procedures for PCRV liner cooling with firewater currently exist (Operations Order 85-17). These procedures will be revised to reflect design changes being made to support interim operation:
-- SLRDIS
-- Spoolpieces to HTFA i l
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! SYSTEMS REQUIRED FOR INTERIM i CORE COOLING USING THE LINElt , [q COOLING SYSTEM WITH FIREWATER j SYSTEM FUNCTION MAJOR ELECTRICAL EQUIPENT HARSH ENVIRONNENT ! SLRDIS/PPS isolate steam ifne break SLRDIS Sensors Yes ; scram reactor and permit SLRDIS Controls No ' building access within Assoc. PPS Controls No one hour. Isolation Valves Yes HE Circulator Isolate reactor coolant Actuating Solenoid Valves Yes Brake & Seal ,
pressure boundary.
. Hydraulic 011 Close hydraulle oft valves.* Central hydraulic package units Yes l
He Puriffcation Depressurization and None N/A Train purification flow path L1 quid Nftrogen LTA cooling during None N/A 5torage depressur1ratfon , Reactor Dutiding Filter gases vented through Exhaust Fans Yes Exhaust Venttistion exhaust stack. Neserve Shutdoom Assure reactivity shutdown None N/A margin. Fire Protection supply firewater to LCS. Fire water pumps No HTFA. HPC, and fuel storage wells. Provide backup cooling for standby diesel generators. t LC5'(Neactor Provide PCRV liner cooling None N/A Plant Cooling flowpath Water System) j 5ervice Water Fireweter outlet floupath. Service water No Cooling for hydraulic cil supply pumps system and standby diesel
; generators.
Cf reviating Water Makeup meter to fire pumps. Circulating Water Makeup Pumps No Electrical Pouer Provide power to the above Standby diesel generators No systems Motor control centers . Yes T* - ., 3 hW hM O N # o r, . m o .. 4
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SELECTION OF FLOWPATHS FOR PCRV DEPRESSURIZATION AND LINER COOLING Existing piping ,' F No automatic valves in flowpath (except with manual overrides) Minimize manual actions Isolate non-essential users
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- Provide redundancy i
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FIRG. WATER ^' RE ACTOR PLT EXHAUST STACK FtREWATER g PUR FICATIOM A MOS. N O c 1301 C-7302 C-13029 Ns )L, M ', _Q h REACTOR PLAuT PRIMARY %E U EYHAUST ' FAus COOLAMT v d E-2305 ,y, g.2305 FROM 'v v v PCRV A 230I E-2301 LOW-TEM PERATORE Ap5ORBERS HIGH TEMPER ATURE , . FILTE R l ADSORDERS . m _ i = l v sy FIREWATER FIREWATER F-13Ol A h h E- 230 4 LOW-TE MPER ATURE y As- 2304 r GAS TO-GAS EXCHAklGER ' j rw )7 :: w
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, F RO M 'v v v A- 1306 PCRV A-250'2.
, E- 2302.
He PURIFICATIOM i COOLERS v- O i F- 2302.
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PURtGLED He FILTERS ]' FI G U RE 1 , PCRV DEPRESGURI 2 ATiou SIMPLIFtED GI.0WPATH l
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