ML20197F654
| ML20197F654 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/07/1986 |
| From: | Hinson C Office of Nuclear Reactor Regulation |
| To: | Lynch O Office of Nuclear Reactor Regulation |
| References | |
| TAC-55287, NUDOCS 8605150640 | |
| Download: ML20197F654 (45) | |
Text
. .
p etray% UNITED STATES g[ ,
p, NUCLEAR REGULATORY COMMISSION
- 1 y ;j WASHINGTON, D. C. 20555 k...+,/ May 7, 1986 _
Docket No. 50-267 ;
MEMORANDUM FOR: Oliver D. T. Lynch, Jr., Section Leader Standardization and Special Projects Directorate Division of PWR Licensing-B FROM: Charles S. Hinson, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
SUBJECT:
SUMMARY
OF MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO DISCUSS 1) FUEL BLOCK CRACKING, 2)
BUILDING 10 CONSTRUCTION, 3) SEISMIC INSTRUMENTATION,
- 4) PCRV TENDONS, AND 5) INTEGRATED LEAK RATE TESTING AT FORT ST. VRAIN (FSV), APRIL 23-24, 1986.
The purpose of this meeting was to discuss the topics of fuel block cracking, Building 10 design and construction, PCRV tendon deterioration, a integrated leak rate testing, and seismic instrumentation at FSV. PSC
. responded to the staff's concerns through formal presentations covering each of the five topics mentioned above. The two NRC reviewers making this trip with FSV Co-Project Manager Charles Hinson were Romuald Lipinski and Robert Rothman, both from the Engineering Branch, DPWRL-B. Other attendees at this l meeting are listed in Enclosure 1.
Plant Tour The licensee conducted the staff on an extensive tour of the plant. The i licensee also performed a lift-off test of a circumferential PCRV tendon for the staff to observe during the tour. Both NRC reviewers had previously indicated their desire to observe a tendon lift-off test to better familiarize themselves with the procedure.
Fuel Block Crackina The staff's main concern on this issue was determining the likelihood of a core overheating accident resulting from a seismic event which would cause cracked fuel blocks to, restrict helium coolant flow through the reactor core.
The licensee addressed the history of cracked fuel blocks at FSV, the dynamic responses of fuel element blocks, and the graphite block surveillance program at FSV. PSC also analyzed the safety impact of a blocked fuel column in which the entire fission product inventory of the fuel column was released. They concluded that the cracking of fuel blocks at FSV would not result in a significant hazard to public health and safety.
PSC provided the staff with three PSC letters (P78146, P78174, and P84104) which further addressed the issues of fuel element cracking and coolant 8605150640 DR 860507 ADOCK 05000267 PDR
- b. .
flow blockage at FSV. The staff agreed to review these letters,las well as the forthcoming Technical Evaluation Report from LANL on the subject, and then provide their evaluation of fuel block cracking at FSV.
Building 10 The staff was interested in the structure-soil-structure interaction effects between the Building 10/ walkover structure and the Turbine building as well as the seismic model used in designing Building 10.
The licensee addressed these issues, as well the two open issues from the staff SER which questioned the adequacy of the gaps between Building 10, the walkover structure,'and the Turbine Building.
The~ licensee agreed to formally submit their responses to the two open issues by May 19, 1986.
Seismic Instrumentation The staff was primarily interested in seeing where the seismic instrumentation and related alarms were physically located within the plant and determining what procedures are to be followed should there be a 1
seismic instrument trigger. This was accomplished during the plant tour.
- The licensee also gave a presentation on seismic instrumentation testing, calibration, and trigger response procedures.
PCRV Tendons The staff was primarily interested in PSC's responses to eleven staff questions on FSV's PCRV tendons.
The licensee addressed these staff questions in detail. PSC also described
- the tendon lift-off test procedure, described the tendon surveillance program, and gave a detailed presentation on observed tendon corrosion, probable causes, and proposed remedial measures to inhibit further tendon corrosion. Enclosure 2 contains the viewgraphs on the FSV prestressing tendon system presented during the mcating.
The staff was not fully satisfied with PSC's responses to the eleven staff questions. The staff will revise these questions based on information presented by PSC during this meeting and submit them formally to PSC at a later date.
PCRV Leakage The staff was primarily interested in the licensee's response to several staff questions concerning FSV's leak rate instrumentation.
The licensee responded to these questions in a presentation which covered methodology for monitoring PCRV leaks, standards / regulations applicable to PCRV leak rate tests, and offsite doses at 35% power for various postulated accidents. Enclosure 3 contains the viewgraphs on the FSV PCRV leakage paths and instrumentation presented during the meeting.
l i'
Overall, the meeting was very informative. The licensee address &d each of the topics thoroughly, although not always to the staff's complete satisfaction. The staff will review PSC's submittals as a means of resolving any remaining staff concerns on these topics.
hub) b > hn Ma-Charles S. Hinson, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
Enclosures:
As stated cc w/ enclosures:
See next page o.
o l Mr. R. F. Walker Public Service Company of Colorado Fort St. Vrair.
cc:
Mr. D. W. Warembourg, Manager Albert J. Hazle, Director o Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. J. W. Gahm, Manager GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. L. W. Singleton, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Fort St. Vrain Nuclear Station Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission
- P. 0. Box 640-P-latteville, 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 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80203 9 %
o Enclosure 1 ATTENDEES AT NRC-PSC MEETING .
APRIL 23-24, 1986 i
Name Organization Charles Hinson NRC-Co-Project Mgr.
R. E. Lipinski NRC-PWR-B/ Engr. Branch Steve Jones PSCO. Results M. N. Holmes PSC Nuclear Licensing Mgr.
Mark Mullen PSC-NED R. L. Craun PSC-Nuclear Site Engr. Mgr.
Bob Gunnerson PSC Super. Structural / Engr.
R. L. Rothman NRC PWR-B/Eng. Branch D. Alberstein GA Technologies L. Milton McBride PSC-Nuclear Fuels and Analysis
, Jim Eggebroten PSC-Technical Services Engr. Supt.
Kathy Owens PSC-Commitment Analyst Don Asmus Stone & Webster-Structural Engr.
Stephen Fisher PSC-Nuc. Fuels Analysis David Goss PSC-Nuclear Licensing Doran Meade PSC-Results Tom Erlevv - PSC-Nuclear Engr.
t . Rich Hell..er PSC-QA
. Tom Moffette PSC-Licensing Consultant Jack Kennedy GA Technologies Tom McIntire PSC-NED Site 4
Enc kcu<e 2 i
i FORT ST VRRlN PCRV.
PRESTRESSING TENDON SYSTEM o GENERRL PCRV PRESTRESSING SYSTEM OVERVIEW o SURVEILLRNCE PROGRAM l
o METRLLURGICRL RSPECTS OF TENDON CbRROSION i
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o TENDON SURVEILLRNCE RESULTS l
l o REMEDIRL MERSURES RND PLRNNED RCTIONS
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i o CROSS HERD 139 B.Sx204 24 % .
I SURVEILLRNCE CRITERIR .
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o CIRCUMFERENTIAL BARREL 15 %
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s i TENDON VISURL INSPECTION PROGRRM i
TOTRL NUMBER TOTRL NUMBER OF TOTRL NUMBER TENDON GROUPS OF NEW TENDONS CONTROL TENDONS OF TENDONS
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n - pO CIRCUHFERENTIRL 13 3 16 TOP CROSS HERD 1 1 2 BOTTOM CROSS HERD 6 2 8 LONGITUDINRL 24 6 30 5
TENDON LIFTOFF PROGRRM TOTRL NUMBER TOTRL NUMBER OF TOTRL NUMBER TENDON GROUPS OF NEN TENDONS CONTROL TENDONS OF TENDONS CIRCUMFERENTIRL 13 3 16 TOP' CROSS HERD 1 1 2 BOTTOM CROSS HERD 3 1 4 LONGITUD 1NRL 12 3 15
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TENDON SURVEILLRNCE SUMMRRI FEBRURRY ~1986 _
PERCENT OF TOTRL TENDON TYPE TOTRL NUMBER LIFTOFF VISURL a h _
LONGITUDINRL 90 82 % 99 %
BOTTOM CROSS HERD 24 83 % 100 %'
. .r TOP CROSS HERD. 24 100 % 100 %
CIRCUMFERENTIRL 310 49 % 55 %
TENDONS WITH RDDITIONRL -
i
, ! NON-EFFECTIVE WIRES -
i NON-EFFECTIVE TENDON DRTE WIRES CONTROL GROUP VM-10 4/84 3 1/85 7 NON-CONTROL GROUP VM-08 4/84 3
. 1/85 4 VM-17 4/84 5 i 1/85 6
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i INTERIM SURVEILLRNCE PROGRRM CONTROL TENDONS NUMBER OF RDDITIONRL NON-TENDON SURVEILLRNCES EFFECTIVE WIRES CM 1.1 2 0 CO 14.4 2 0 i CM.16.3 2 0 TIRM2 2 0 BIRM4 4 0 BILM3 4 O VM-10 4 4 VI-20 4 0 VM-20 4 0 VM-37 4 0 VI-40 4 0 VM-40 4 0 i
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TENDON LORD LONGITUDINRL TENDONS (KIPS) /
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l MINIMUM PREDICTED PRESTRESS LOSSES 1400 =_ _ _ _ _ __ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,
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1200 - LohD CELL INDICRTED RANGE -
LORD CELL LOH-L 6EL RLRRH SET POINT -
1100 -
MINIMUM DESIGN LORD 1000 -
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~ LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
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REVIEW 0F l STANDARDS AND REGULATIONS APPLICABLE TO FORT ST, VRAIN PCRV HELIUM LEAKAGE RATES
- - AND TESTING
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PRESSURIZING SYSTEM TECH SPEC COMPARISON WESTINGHOUSE STANDARD TECH SPEC (WSTS) 3.6.1.1 -
VS. FSV TECH SPEC UPGRADE (TSUP) 3.6.1.3 i
WSTS 3.6.1.1 CONTAltfENT ISOLATION VALVE AND CHANNEL WELD FHt55URllAIION SYSItfl5 SYSTEM SHALL BE OPERABLE TSUP 3.6.1.3 PCRV - INTERSPACE MINIftN PRESSURIZAlION INTERSPACES BETWEEN PRIMARY AND SECONDARY PCRV CLOSURES SHALL BE PRESSURIZED WITH PURIFIED HELIUM:
A. GREATER THAN PRIMARY COOLANT PRESSURE, OR
. B. WITH A REHEATER LEAK, STEAM GENERATOR INTERSPACES MAY BE LESS THAN PCRV PRESSURE, BUT GREATER THAN REHEAT STEAM PRESSURE.
i COMPARISON - FSV TSUP EQUIVALENT TO WSTS
(.
ACTION WSTS - RESTORE WITHIN 7 DAYS, OR HOT STANDBY WITHIN NEXT 6 HOURS TSUP - RESTORE WITHIN 24 HOURS, OR SHUTDOWN WITHIN HEXT 24 HOURS COMPARISON - FSV TSUP MORE RESTRICTIVE THAN WSTS SURVEILLANCE REQUIREENTS
~ '
WSTS 4.6.1.4 - VERIFY PRESSURIZED ONCE PER 31 DAYS TSUP 4.6.1.3 - VERIFY WITHIN PRESSURE LIMITS ONCE EVERY 24 HOURS COMPARISON - FSV TSUP MORE CLOSELY MONITORED THAN WSTS CONCLUSIONS ON PRESSURIZING SYSTEM TECH SPECS FSV TSUP IS OVERALL MORE CLOSELY MONITORED AND MORE RESTRICTIVE THAN WSTS.
LEAKAGE TECH SPEC C0FPARIS0N Q
WESTINGHOUSE ST/JOARD TECH SPEC (WSTS) 3.6.1.2 VS. FSV TECH SPEC UPGRADE (TSUP) 3.6.1.4
~
WSTS 3.6.1.2 CONTAlW ENT LEAKAGE A. OVERALLI_NTEGRATEDLEAKAGERATE{LAATPA,OR{LTATPr I B. FENETRATIONS AND VALVES 5; 0.6 LA AT PA MiERE IN ACCCRBANCE WITH 10CFR50 APPENDIX J:
PA'= CALCULATED PEAK. ACCIDENT PRESSURE La = TECH SPEC ALLOWABLE LEAK RATE AT PA PT = REDUCED TEST PRESSURE FOR PERIODIC TESTS LT = MAXIttJM ALLOWABLE LEAK RATE AT Pr TSUP 3.6.1.4 PCRV CLOSURE LEAKAGE A. PRIMARY CLOSURES -
EQUIVALENT TO 400 LB/ DAY / GROUP = 16.7 LB/HR 'LT) AT 10 PSID (PT).
BASED ON ASSLEING SECONDAP.Y CLOSURE RUPTURE:
- PA = 688 PSID ACROSS PRIMARY CLOSURE
- LA = 1145 LB/HR.THROUGH PRIMARY CLQSURE GROUP
- ACTIvlTY RELEASE EQUIVALENT TO ITA
- - EAB DOSE CONSERVATIVELY CALCULATED:
ASSUMING "DE$1GN" ACTIVITY-IN COOLANT ASSUMING X/0 = 2.7 x 10-3 SEC/M3:
( = (38 x MCA)
= (7 x DBA-2)
EAB DOSE IS ORDER-OF-MAGNITUDE BELOW 10CFR100.
B. SECONDARY CLOSURES TOTAL - EQUIVALENT TO 400 LB/ DAY AT 688 '
PSID (ECONOMIC LIMIT - PURIFIED HELitE).
C. STE M GENERATOR PENETRATIONS WITH LEAK TO REHEAT SYSTEMS:
700 LB/ DAY / GROUP PURIFIED HELItE, BASED ON CONDENSER VACl?JM LIMIT (TSUP LC0 3.6.1.5) 200 CPM GROSS ACTIVITY IN STEAM GENERATOR INTERSPACE GROUP. (200 CPM IS APPROX 2X BACKGROUND) l (CURRENT LC0 4.2.9 - BEING TRANSFERRED TO SPEC 8.0,
" RADIOLOGICAL AND ENVIRONMENTAL TECHNICAL SPECIFICATIONS.") _ LIMIT OF 1.4 CURIE / DAY THROUGH l
PRIMARY' CLOSURES EASURED AT AIR EJECTOR EXHAUST.
BASED ON 10% OF PLANT'S DESIGN OBJECTIVE GASEOUS RADI0 ACTIVITY RELEASE OF 4160 CURIES / YEAR Ato 80% PLANT l CAPACITY FACTOR.
l 1.4 CI/ DAY =
0.6 LBS/ DAY PRIMARY HELIUM AT " DESIGN" ACTIVITY = 60 LBS/ DAY AT " ACTUAL" ACTIVITY COMPARISON -
FSV HAS MORE DIVERSE RESTRICTIONS THAN Wsis (CONTINUED)
LEAKAGE TECH SPEC COMPARISON (CONTINUED)
ACTION ,
6 WSTS - REDUCE LEAK RATE BEFORE EXCEEDING 200 DEGREES F TSUP - REDUCE LEAK RATE WITHIN 24 HOURS OR BE IN SHUTDOWN WITHIN NEXT 24 HOURS COMPARISON -
DIFFERENCE IN ACTIONS BECAUSE FSV CONTINUOUSLY MONITORS AND FREQUENTLY TESTS DURING OPERATION; PWRS INFREQUENTLY TEST WHILE SHUTDOWN.
SURVEILLANCE REQUIREE NTS WSTS 4.6.1.2.G - EXEMPTS PRESSURIZED PENETRATIONS CONTINUOUSLY
. MONITORED FROM TYPE B PERIODIC TESTS, PROVIDED PRESSURIZING SYSTEMS ARE OPERABLE PER STS 4.6.1.4.
WSTS 4.6.1.2.1 - REQUIRES TYPE B TESTS FOR EVERY 3 YEARS FOR PENETRATIONS CONTINUOUSLY MONITORED FOR LEAKAGE TSUP 4.6.1.4 - ONCE PER 92 DAYS, OR WITHIN 24 HOURS AFTER AN UNANTICIPATED INCREASE IN PRESSURIZATION GAS FLOW, DETERMINE WE LEAK RATE OF:
(A) PRIMARY CLOSURES (B) SECONDARY CLOSURES (C) STEAM GENERATOR INTERSPACE-TO-REHEAT SYSTEM TSUP 4.6.1.5 - ONCE PER 12 HOURS, CHECK ACTIVITY MONITOR IN STEAM GENERATOR INTERSPACE WITH PRESSURE BELOW PRIMARY PRESSURE (I.E., WITH REHEATER LEAK)
COMPARISON - FSV MORE RESTRICTIVE THAN WSTS CONCLUSIONS ON LEAKAGE AND PRESSURIZING TECH SPECS FSV IS MORE CLOSELY MONITORED AND OVERALL MORE RESTRICTIVE THAN WSTS. ,
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10CFR50, APPENDIX J PRIMARY REACTOR CONTAlffENT LEAKAGE -
TESTING FOR WATER-COOLED POWER REACTORS FEBRUARY 14,1973 ;
INTRODUCTION 10CFR50.54(0) SPECIFIES WATER-COOLED POWER REACTORS SHALL COMPLY WITH APPENDIX J. (NOT REQUIREMENT, BUT l-SV MEETS INTENT.)
(A) PURPOSE OF TESTS ASSURE LEAKAGE SHA NOT EXCEED ALLOWABLE LEAKAGE RATE VALUES AS SPECIFIED IN THE l(LECHNICAL (FSV MEETS INTENT.)
(B) PERIODIC SURVEILLANCE OF REACTOR CONTAlft1ENT PENETRATIONS AND ISOLATION VALVES IS PERFORfED SO THAT PROPER MAINTENANCE AND REPAIRS ARE MADE DURING THE SERVICE LIFE. (FSV MEETS INTENT.)
LEAKAGE TESTING REQUIRENNTS TYPE A TESTS (0VERALL INTEGRATED LEAKAGE RATE. APPENDIX J REFERENCES
,. ANSI N W 4 - 1972)
L r . PREOPERATIONAL -
FSV PERFORMED IN 1971 (REFS. 1 AND 2) IN
~ACCORDANCE WITH GA TEST SPEC. ll-X-14, ISSUE E, OCTOBER 11, i 1971, "PCRV LEAK TEST PROCEDURE", WHICH APEARS VIRTUALLY C. SAME AS ANSI N 45.4-1972, (APPARENTLY BASED ON DRAFT OF N 45.4).
PERIODIC RETESTS - FSV PCRV AND PENETRATIONS CONTINUOUSLY PRESSURIZED AND RATE OF LOSS CONTINU0USLY MEASURED. -
l
! REFERENCE 1: GULF-GA-A10875 " FORT ST. VRAIN UNIT 1 PCRV LEAK TEST I REPORT FOR PUBLIC SERVICE COMPANY OF COLORAD0", BY R.
BARKER, U.K. RATH, AND C.E. TABNADGE, GULF GENERAL ATOMIC CO., FEBRUARY 1, 1972
(
REFERENCE 2: GA-A13383 (CONF-150906--2) "LEAKTIGHTNESS IN HTGRS -
EXPERIENCE AT FORT ST. VRAIN," BY A.J. NEYLAN, R.A.
BARKER, AND A.F. DEARDORFF, GENERAL ATOMIC, JUNE 10, i 1975. PRESENTED AT INTERNATIONAL CONFERENCE ON l EXPERIENCE IN THE DESIGN, CONSTRUCTION, AND OPERATION OF PRESTRESSED CONCRETE REACTOR VESSELS AND CONTAlft1ENTS FOR NUCLEAR REACTORS, SEPTEMBER 8-12, 1975, YORK, ENGLAND.
(CONTINUED)
10CFR, APPENDIX J (CONTINUED)
TYPE B TESTS (LOCAL TESTS OF PENETRATION PRESSURE-C0t AINING OR LEAKAGE-LIMITING BOUNDARIES)
PREOPERATIONAL - FSV PERFORMED IN 1971, IN ACCORDANCE WITH GA TEST SPEC. ll-X-16, "PCRV PENETRATIONS - PROOF PRESSURE AND LEAKAGE TESTS". USED RATE-OF-PRESSURE-LOSS E THOD.
PERIODIC - REQUIRED EVERY 2 YEARS BY APPENDIX J.
FSV TECH SPEC SURVEILLANCE REQUIREMENT 5.2.16 REQUIRES PENETRATION LEAK TESTS ONCE EACH QUARTER.
FSV SURVEILLANCE TEST SPECIFICATION SR 5.2.16 A-0, "PCRV CLOSURE LEAKAGE DETERMINATION", SPECIFIES MEASUREMENT OF FLOW RATE. IF LEAK RATE APPEARS LARGE, IT IS RETESTED WIm PROCEDURE SR-E-151-X, " PENETRATION INTERSPACE LEAKAGE PRESSURE DECAY TEST", WHICH USES
- PRESSURE-DECAY ME m0D B AT IS MORE EXACT W AN FLOW-RATE METHOD.
l CONTINUOUS-(PERMITTED BY APPENDIX J - NOT REQUIRED)
ALL FSV PENETRATIONS CONTINUOUSLY PRESSURIZED WITH PURIFIED HELIUM AND FLOW-RATE ALARMED. (TS LCO 4.2.7,
( SR 5.2.15)
TYPE C TESTS - (NA TO FSV)
FSV HAS NO VALVES FITTING APPENDIX J DEFINITION OF TYPE C VALVES, -
I.E., W AT CONNECT DIRECTLY TO ATMOSPHERE, EXCEPT SAFETY VALVES WHICH ARE PROTECTED BY RUPTURE DISKS. LEAKAGE BEWEEN RUPTURE DISKS AND SAFETY VALVES LIMITED AND MONITORED (TS LC0 4.2.7, SR 5.2.1).
' " OPERABILITY OF FSV PRESSURIZATION SYSTEM ISOLATION VALVES COVERED BY TECHNICAL SPECIFICATION ON SURVEILLANCES (SR 5.2.16).
CCNCLUSIONS EGARDING FORT ST VRAIN NOT A REQUIREMENT OF GAS-COOLED REACTORS, BUT FSV MEETS INTENT.
NOTE: APPENDIX J REFERENCES ANSI N 4.5.4 (1972), BUT DOES NOT REFERENCE ANSI /ANS-56.8-1981.
1 l
ANSI N 45.4 (1972)
~
" LEAKAGE RATE TEST OF CONTAltf0IT STRUCTURES FOR NUCLEAR REACTORS" ;
- APPROVED MARCH 16,1972 BY ANSI PURPOSE AND SCOPE: SPECIFIES UNIFORM METHOD FOR DETERMINING THE (INTEGRAL) LEAKAGE RATE OF CONTAlffENT STRUCTURES.
i TEST %THODS: EITHER ABSOLUTE OR REFERENCE-VESSEL ETHOD PERMITTED.
PROCEDURES FOR BOTH ETHODS ARE DESCRIBED.
APPLICATION TO FORT ST. VRAIN L
.h . INTEGRAL LiAK TEST OF PCRV PERFORED IN 1971, PRIOR TO APPROVAL OF A
- STANDARD BY ANSI. FSV TEST FOLLOWED ETHODS OF DRAFT ANS 45.4.
l ~
INTEGRAL LEAK RATE AT 688 PSIG DETERMINED BY:
(A) ABSOLUTE METHOD CM -
BETWEEN 0 AND 14.38%/YR FOR MOLECULAR FLOW BETWEEN 0 AND 5.64%/YR FOR VISCOUS FLOW (B) REFERENCE VESSEL ETHOD BETWEEN 0 AND 8.84%/YR FOR MOLECULAR FLOW i -
BETWEEN 0 AND 3.47/YR FOR VISCOUS FLOW MORE PROBABLE VALUE WAS 3.47%/YR BASED ON REFERENCE VESSEL E THOD FOR VISCOUS FLOW, WHICH IS A LEAK TIGHTNESS EQUAL TO THE INSTRt#ENT ERROR. (REFERENCES 1 AND 2)
REFERENCE 1: GULF-GA-A10875 " FORT ST. VRAIN UNIT 1 PCRV LEAK TEST REPORT FOR PUBLIC SERVICE COMPANY OF COLORAD0", BY R.
BARKER, U . K. RATH, AND C.E. TAPNADGE, GULF GENERAL ATOMIC CO., FEBRUARY 1, 1982.
REFERENCE 2: GA-A13383 (C0W-150906-2) "LEAKTIGHTNESS IN HTGRs -
EXPERIENCE AT FORT ST. VRAIN", BY A.J. NEYLAN, R.A.
BARKER, AND A.F. DEARDORFF, GENERAL ATOMIC, JUNE 10, 1975. PRESENTED AT INTERNATIONAL CONFERENCE ON EXPERIENCE 'IN THE DESIGN, CONSTRUCTION, AND OPERATION 0F PRESTRESSED CONCRETE REACTOR VBSSEL AND CONTAINMENTS FOR NUCLEAR REACTORS, SEPTEMBER 8-12, 1975, YORK, ENGLAND.
1
ANSI /ANS - 56.8 - 1981 "CONTAINMFN SYSTEM LEAKAGE TESTING REQUIREMENTS"
~
APPROVED FEBRUARY 19,1981 BY ANSI PURPOSE AND SCOPE: I PROVIDES BASIS FOR DETERMINING LEAKAGE RATES THROUGH THE CONTAINMEf6 SYSTEM OF WATER COOLED REACTORS. SPECIFIES MINIMUM LEAKAGE RATE TEST REQUIREMENTS.
NOTE 1: ANSI /ANS 56.8 IS NOT REFERENCED BY 10CFR50 APPENDIX J, BUT APPARENTLY IS WRITTEN AS AN OPTIONAL STANDARD MEANS OF IMPLEMENTING APPENDIX J REQUIREMEfRS. APPARENTLY 5.6.8 IS NOT YET A REQUIREMENT FOR LWPs.
CONJUNCTIVE CODES AND STANDARDS:
MENTIONS ONLY N41.51 (IEEE STANDARD 317-1976), " ELECTRICAL PENETRATIONS IN CONTAINMENT STRUCT1JRES".
NOTE 2: THERE IS NO MENTION OF ANSI N 45.4, WHICH ANSI /ANS 56.8 APPARENTLY IS INTENDED TO REPLACE OR SUPPLEENT.
- LEAKAGE TESTING REQUIREFENTS TYPE A TESTS PREOPERATIONAL - FSV 1971 TEST PERFORMED TO DRAFT OF ANSI N D.4 APPEARS TO MEET ANSI /ANS - 56.8.
TYPE B TESTS PRESSURE-DECAY AND FLOW-RATE METHODS ARE ACCEPTABLE TO ANSI /ANS 56.8. FSV USES BOTH METHODS.
CONCLUSIONS REGARDING FSV NOT APPLICABLE TO GAS-COOLED REACTORS, Bur FSV MEETS IfREtR.
l t
=. . _ . . . . _ . . -- -. _. _ . ._
-s -
USNRC STANDARD REVIEW PLAN 6.2.6 "CONTAllfENT LEAK TESTING" REv. 2 - JULY 1981 i '
I. AREAS OF REVIEW: CONTAIENT LEAKAGE TESTING PROGRAM REVIEWED i
FOR CONFORMANCE TO 10CFR50 APPENDIX J AND GENERAL DESIGN CRITERIA 52, 53, AND 54*.
NRC REVIEW OF 'CONTAINE NT LEAKAGE TESTING PROGRAM COVERS SPECIFICALLY:
- 1. INTEGRATED LEAKAGE RATE TESTS (APP. J, TYPE A)
- 2. PENETRATION LEAKAGE RATE TESTR (APP. J, TYPE B)
- 3. CONTAI E NT ISOLATION VALVE TESTS (APP. J, TYPE C)
- 4. TECH SPECS ON LEAKAGE RATE TESTING
- 11. ACCEPTANCE CRITERIA: CONTAI E NT LEAKAGE RATE TESTING PROGRAM, AS DESCRIBED IN SAR, IS ACCEPTABLE IF IT E ETS APPENDIX J.
APPENDIX J IS ACCEPTABLE BASIS FOR SATISFYING GDC 52, 53, AND 54.*
i .
EXCEPTIONS TO APPENDIX J REVIEWED ON CASE-BY-CASE BASIS.
{} APPLICATION TO FORT ST. VRAIN:
FSAR: THE FSAR CONTAINS INFORMATION E ETING SRP 6.2.6 REQUIREMENTS 1
FOR HELItE LEAKAGE TESTING INFORMATION IN SECTIONS:
' 5.8.2 13 LEAKAGE (DESIGN OF PENETRATIONS AND CLOSURES) .
5.12.2 .EAKAGE (LIMITING CONDITIONS FOR OPERATION) 5.13 MEOPERATIONAL LEAKAGE TESTS OF PCRV, 5.13 1.2',Y
.2.3 NENETRATIONS, LINERS, AND CLOSURES.
5.13.9 3URVEILLANCE OF PCRV PENETRATION INTERSPACE PRESSURE INSTR M NTATION 5.13.10 SURVEILLANCE OF PCRV CLOSURE LEAKAGE
- 5.13.11 ' INSPECTION OF PCRV PENETRATION AND CLOSURES 7.3.5.1 REHEAT STEAM RADIATION MONITORING i L
FSV FSAR MEETS INTENT OF SRP 6.2.6.
_lL2tU IN COMPLIANCE WITH EARLIER GDCS - SEE FSAR APPENDIX C, GDCS 53, 54, 55, 56,AND 57, WHICH COVER SAME GENERAL MATERIAL AS CURRENT GDCS 52, 53, AND 54.
FSV EETS THE LETTER AND/0R INTENT OF GDCS, APPENDIX J: FORT ST. VRAIN M ETS INTENT OF 10CFR50 APPENDIX J (SEE EARLIER DISCUSSION).
(CONTINUED)
. ~ . , - - . _ - - .
.-,..-c.--.- - . -
CONCLUSIONS REGARDING STANDARD REVIEW PLAN 6.2,6:
I ALTHOUGH, FORT ST. VRAIN FSAR AND TECH SPECS WERE REVIEWED AND ACCEPTED AND OPERATING LICENSE WAS GRANTED BY AEC PRIOR TO ISSUANCE OF STANDARD REVIEW PLAN, AND ALTHOUGH COWLIANCE WITH STANDARD REVIEW PLANS IS NOT REQUIRED BY NRC,-
- FORT ST. VRAIN MEETS THE INTENT OF SRP 6.2.6.
5 e
e o?*
~
)
REVIEW OF
~
TABLE 2-1 (ON HELIUM LEAK RATES) I 0F ArrAcif1ENT 1 AND i
TABLE 1
SUMMARY
OF 0FF-SITE DOSES
, RESULTING FROM POSTULATED ACCIDENTS" 0F ATTAC MENT 5 i -
{s TO LETTER P-85460
SUBJECT:
CONFIRMATORY ACTIONS IN SUPPORT OF 35% POWER RESTRICTION DURING EQ
- , ,y SCHEDULE EXTENSION PERIOD i
+ /1.1./ TG l
Attachment 1
- to P-85460 TABLE 2-1 i
NORMAL CONDITIONS ALLOWABLE DIFFERENTIAL ACTUAL (1) MAXIMUM (2)
LEAK PATH PRESSURE LEAK RATE LEAK RATE Primary Closures
- All except S/G 10 psi 31.2 lb/ day 400 lb/ day penetrations with (GroupVII) per Group CRH leak
- S/G penetrations with CRH leak
- at actual circu- 350 psi (3) 60 lb/ day total lating activity maximum
- at design circu- 350 psi (3) 0.6 lb/ day l -
lating activity maximum total Secondary Closures 688 psi 3.89xE-4 400 lb/ day
('/
L lb/ day total Cold Reheat Pipe 15 psi 31.2 lb/ day 700 lb/ day l in Loop with Cold (4).
Reheat leak .
t pw/
.e f -r.5.
i e
es-
-.----_...-.---m- _ . . . . , - _ - . , . , . - . ~ . . , . . . . , - - _ _ . - , - . _ - - . . _ . . - _ . _ _ , . _ . - . . , _ _ . - - - - _ _ ~ - - - , . - - - -
Attachment 1 to P-85460 _
NOTES TO TABLE 2-1 I Notes: (1) Reference FSV Surveillance Test Procedure SR 5.2.16a-Q, PCRV Closure Leakage Determination, performed November 7-8, 1985 (2) Maximum Allowable leak Rate pennitted by Technical Specifications, LC0 4.2.9 - PCRV Closure Leakage (3) S/G penetration primary closure leakage cannot be ~
measured directly or indirectly. Leak tightness is inferred from absence of activity in the interspace.
H (4) Measured leakage in Group M is attributed to the '7"yp e CRH piping internal to the penetration. This is purified helium leakage at this time.
i 1
o
. eV y- e t we-- -w- - --e-+--*n - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- -- -
m----- v-n-'
. Attachment 1 to P-85460 l A C TUA L L E. A K A C E per svevWito5ee T'est s it S.I. !' . a-Q N ov+mber M~9s I98 PRIMARY SECONDARY .
CLOSURE CLOSURE GROUP NO. PENETRATIONS LEAKAGE LEAKAGE I All penetrations in the 0 0 top head of the PCRV (37-control rod drive, 2-HTFA, and 1-top access)
II All instrument penetra- 0 0 tions (20), plus the bottom access penetration
.- III Total of 6-Loop I Steam 0 0 Generator Penetrations
' ~
IV Total of 6-Loop II Steam 1.3 lb/hr* = 31,1 lb/ day Generator Penetrations i
- V Helium Circulator 0 0 r -
VI Helium Circulator 0 0 h VII Heliu'm Circulator 1.3 lb/hr 1.62xE-5 = 3.8 9 x E- 4 lb/hr Ib/ d *y VIII Helium Circulator 0 0
- There are no provisions for measuring Group III or IV secondary seal leakage, so only the total primary plus secondary leakage plus leakage to the cold reheat (if any) can be determined. This leakage is assumed to be cold reheat leakage based on the absence of activity in the interspace the presence of helium in the main
~ -'
condensor and the leak tightness of all other secondary closures.
Alsa - s e conda.y e. tos v><s om s Frem - ym%-
('tke+rabicht etyf 1enl~ w elded,
! - 9-
/
w I
- - ~ , - - , - , , - - ~ , , , . , - , - , , - , - - - - , - - , - , - - - - - - - - - - - - - - - , , - , - - - - - - . , . - . . - , - - - - - , - - - - - - - - - - - - - - - . . ,
'q g . . .
TABLc'2-1 .. Attachment 1 (Supplemented 4/13/86) to P-85460' .
PCRV PENETRATION LEAKAGE NORMAL CONDITIONS ALLOWABLE BASIS FOR DIFFERENTIAL ACTUAL (1) MAXIMUM (2) ALLOWABLE Path PRESSURE LEAK RATE LEAK RATE LEAK RATE No. LEAK PATH ( A PN) PER LCO 4.2.9 I Primary Closures: "
! 1A. -- All except S/G 10 psi 31.2 lb/ day 400 lb/ day '
1145 lb/hr at 688 psi if penetrations with (Group VII) per Group secondary closure fails
- CRH leak (A) (A) ;
LB. -
S/G penetrations with I 1.4 curie / day:
CRH 1eak: (Group IV)
I.B.1 -
at actual circulating 350 psi (3) (4) 60 lb/ day activity Estimated from " Actual" maximum (B) none total Activity (100 X Design)
I.B.2 -
at design circulating 350 psi (3) (4) 0.6 lb/ day activity Calculated from " Design" maximum (B) none total Noble Gas Activity per Note C II Secondary Closures 688 psi 3.89 x E-4 400 lb/ day Purified Helium Economic (Total) _
lb/ day total Limit Group VII III Cold Reheat Pipe in Loop 15 psi 31.2 lb/ day 1700 lb/davl If primary closure leaks with Cold Reheat Leak (B) (4) Purified (if radioactive), Path IB (Group IV) Helium limiting Nat2s: '
(A) Solid boxes indicate directly ' rom LCO 4.2.9 Dashed boxes indicate directly from LC0 4.2.9 " asis 4
(B) Estimated maximum differential pressures (C) 0.6 lb/ day calculated from 1.4 Ci/ day LCO limit:
3.4 Ci l y ~ Primary Coolant He inventory (FSAR Sect. 4.2.1) '
4 4 day
.3
" Design" primary coolant Noble Gas Activity (Table 3.7-11
=
1.4 Ci "y 7370 # He . 0.648 fHe
. day .15.922 Ci Noble Gases.
day M
~
i
~
Attachment 1 to P-85460 ,
TABLE 2-1 (continued) I ACCIDENT CONDITIONS LEAK RATE BASED ON ALLOWA8LE DIFFERENTIAL ACTUAL MAXIMUM LEAK PATH PRESSURE LEAK RATE LEAK RATE Primary Closures
- All except S/G 688 psi 10.8 lb/hr 1145 lb/hr penetrations with maximum (GroupVII) per Group CRH leak
- S/G penetrations with CRH leak
- at actual circu- 688 psi 0 <100 lb/ day l lating activity maximum total
- at design circu- 688 psi 0 <1.0 lb/ day lating activity maximum total Secondary Closures 688 psi 3.89xE-4 400 lb/ day maximum lb/ day total Cold Reheat Pipe 688 psi 211 lb/ day 4740 lb/ day .
maximum
\
M&) f<
s.,,a e x sd =c,
g p '- ,
TABLP2-1. Attechment 1 (Supplemented 4/13/86) to P-85460 .
j PCRV PENETRATION ' LEAKAGE ACCIDENT CONDITIONS i (AP) A LEAK RATE BASED ON BASED ON
' ALLOWABLE PATH DIFFERENTIAL ACTUAL
' MAXImlM COMENTS AND APPLICATION NO. LEAK PATH PRESSURE LEAK RATE LEAK RATE TO DOSE CALCULATIONS I. Primary Closures:
1.A. - All except S/G 688 psi 8
'30.8 lb/hr Limited by Path II to penetrations with [naximunj (Group VII) l1145 lb/hh g g Grouy g 400 lb/ day CRH Leak (all ex-cept Group IV) (A) (D) (A)
- I.B. - S/G penetrations with j CRH leak: (Group IV)
! 688 psi 0 <100 lb/ day total Not used because Dose Cales.
1.B.1 - at " actual" circulating maximum (D) based on " Design" Activity.
activity (E & F) l (Path I.B.2) l I.B.2 - at " Design" circulating 688 psi 0 <1.0 lb/ day total Basis for " Design" Activity
! activity maximum (D) Dose Calcs (E & F)
II. Secondary Closures 1688 nsil 3.89 X E-4 1400 lb/ day totall Basis for Dose Calcs
- (total) maximum lb/ day (D) l III. Cold Reheat Pipe 688 psi 211 lb/ day 4740 lb/ day Limited by Path II.B to i- (Group IV) maximum (D) (D) I to 100 lb/ day (F & G)
[
! Notes: (D) Accident Leak Rate (LR A ) assumed extrapolated proportional to square root of pressure: **
(E) Penetration Interspace Pressurization assumed failed and flow resistance of reheater leak assumed negligible.
j (F) HELBA assumed to be cold reheat pipe break (non-isolable).
j (G) Flow resistance of primary closures assumed negligible.
i
PCRV PENETRATION LEAKAGE ASSUMPTIONS
~
(TABLE 2-1 " NORMAL" AND " ACCIDENT")
i ,
- 1. CONSERVATIVELY BASED ON TECH SPEC ALLOWABLES (" ACTUAL" LEAK RATES MUCH SMALLER).
- 2. SECONDARY CLOSURES MORE LIMITING THAN PRIMARY CLOSURES (EXCEPT STEAM GENERATOR PENETRATION WITH REHE/iER LEAK).
USE 400 LB/ DAY (AT 688 PSID) AS BASIS FOR DOSE CALCULATIONS, (PLUS REHEATER LEAK).
- 3. REHEATER LEAK LIMITED BY TECH SPECS TO 1.4 CURIE / DAY NORMALLY.
= 0.6 LB/ DAY " DESIGN" PRIMARY HELIUM
=
60 LB/ DAY " ACTUAL" PRIMARY HELIUM AB0VE LEAKAGES LIMITED BY PRIMARY CLOSURES.
PRIMARY CLOSURE DIFFERENTIAL PRESSURE 350 PSI AT 35% POWER l IN STEAM GENERATOR PENETRATIONS WITH REHEATER LEAKS.
- 4. ASSUE REHEAT PRESSURE 0 PSIG AFTER ACCIDENT. ASSUE PENETRATION INTERSPACE PRESSURE IS VERY LOW.
PRIMARY CLOSURE DIFFERENTIAL PRESSURE 688 PSID MAXIMJM AFTER ACCIDENT.
- 5. STEAM GENERATOR PENETRATION PRIMARY CLOSURE ACCIDENT LEAK RATE *
- CALCULATED.
" DESIGN" 0.6 1 X DAY PSI SW PSI) N
) / 1.0.JL)AY
~
" ACTUAL" 60 1 X b66 PSID 100 1 DAY $W PSID DAY
- 6. MAXIKIM OFFSITE DOSE BASED ON " DESIGN" ACTIVITY. THEREFORE,USE 1.0 LB/ DAY REHEATER LEAK AS BASIS FOR DOSE CALCULATIONS.
- 7. TOTAL LEAK:
REHEATER LEAK (1 LB/ DAY) X (2 HR) =
0.083 LB SECONDARY CLOSURES (400 LB/ DAY) X (2 HR) = 33.3 LB CIRCULATOR SHAFT SEALS (60 MIN TO SEAL) 1250. LB TWO HOUR TOTAL 1283.383 LB
I CONCLUSIONS REGARDING HELIUM LEAKAGECALCULATIONS CIRCULATOR SHAFT SEAL LEAKAGE IS DOMINANT r
TEREFORE - E THOD OF EXTRAPOLATION OF REHEATER LEAKAGE AT DIFFERENT
- PRESSURES IS NOT SIGNIFICANT.
L
.~ ,
O e eV l
l
Attachment 1 to P-85.460 De e 10, 8 9 Ff TABLE 2-2 I DetEt F IL O M P NE TA AT cN AND CIR ev L A tor S H A PT SEAL LEAKS- d miwter to set s ka f t sea ()
INTEGRATED DO5E i; REM)
LOCATION TIME WHOLE BODY THYROID Exclusion 2 Hours 5.4 x E-3 2.3 x E-2 Area Boundary Low Popu- 30 Days J g x E-4 M x E-2 :K lation g,g 3,2 Zone Boundary
.* L P 1 dos e s re vis ed clo m ard is 8.e v, A i
of @A ry ort 109613 J (619,19 76 O
fee TABLE 1 Of 6A efport 9 00 6 9 3/A for a s s uwsp t ion s for cal e vie 4 '~ow e of doses , -
, . _ . , - . - .,s- - _ - - - - - -
,- c- - - - - - - - - - - - - - - - - - - - - -
CA PePort i 908693/A Jew is, tWG TABLE 1 ASSUMPTI'ONS FOR CALCULATIONS OF OFFSITE AND PERSONNEL DOSES From l%e% han and Cimvlak $l'*N S'*I Leaks Parameter .. Assumption SOURCE TERMS C ons M a've Gas-Borne FSV FSAR Table 3 7-1 Amendment 1E Mo' "879 MW(t) Design" Primary Coolant *
. System Activities (IOS% P*w'*) Jf4Wer E
~ RELEASE RATE TO A1HOSPHERE FROM 505/ Day
- REACTOR BUILDING %
, FILTRATION % None EXCLUSION AREA BOUNDARY 590 m LCW POPULATICN ZONE 16000 m -
WIND SPEED 1.0 m/s BREATHING RATE 3 47 x 10-* m8 /s
" ATMOSPHERIC DILUTION FACTORS M FSV FSAR Table 14.12-1 jd=, as Bouyancy Neglected: Cr**ad live 4.0 x 10-* s/m8 9 EAB DBh-Lj og{ye 3 18 x 10-8 s/m8 i LPZ Rep.J Derre33-vregobon DECAY AND BUILDUP ENROUTE Considered ACC4'"'
- A ss umiw9 Re. e.kr Bu.ldig WJ,-( pon Ek usf l=a n sad H lt
- r Sy & m not op aHosat Pa6e 3
i UPDATED FSAR Revision 3 D6A-2 Kap.J Dynnven4 %
Table 14.12-1 8
SHORT TERM DILUTION FACTORS (sec/m )
. FOR WORST ATMOSPHERIC STABILITY CONDITION ANO 1 m/s With Helium Buoyancy '
Distance (meters) Particulate ithout Helium Gaseous Buoyancy
, 590 9.56E-06 (A) 9.34E-06 (A) 4.0E-04(G)*
16000 7.75E-06 (F) 1.98E-06 (E) 8E-05 (G)*
- Based'on Table XXX of Revised Supplemental Climatological Report, Nov. 1970, FSAR Appendix G.2.
~
l TABLE 1 atta:nten: E
~
SUMMARY
OF 0FF-SITE DOSES RESULTING FROM POSTULATED ACCIDENTS 00
, , o, q y 7hl
.[ . 1 TOTAL DURATION 005E (REM) .go og c. E l LOCATION OF '
ACCIDENT MAXIMUM DOSE WHOLE BODY THYROID .
Complete Loss of Low Population 3.7 x E-4 3.6 x E-2 FSAR Tate 14. l 3-t i Forced Circulation Zone Boundary Cooling - DBA No. 1 ,j l8gdy)
Worst PCRV Penetra- Exclusion Area 2.5 17.4 '
tion Failure (both Boundary O #d f d ro6te Ph n-1 closures of a steam Dilu Wo n -Tab te E L-l (2 Hours) a generator penetra- 5 9 o -4e 3 ground tion) - DBA No. 2 ("*/((E'"leverdM 01.Bg'd.y 1/dt a 4.0 4; E%'$ t S
" Maximum Credible Dates-PsAR "Ikble 14.13-l Exclusion Area 1.62 x E-1 8.8 x E-2 bilvhen -
Accident" (largest Boundary 0**' d 8" "^"'t Potential PCRV leak (2 Hours) %,$'rD %*" #
rate) S9 o **4 rr 1/d e 7x E-5'see/d 30 minutes to set Low Population 4 x E-4 2.1 x E-2 circulator seals and Zone Boundary g yf,,,g.gogf13A 400 lb akage via (30 day)
. PCRV p n ration Offritebres'TeNe3 closures Exclusion Area 3.6 x E-3 1.5 x E-2 Arevmp+/en For
- *4ff lkf + d */4 7 Boundary c,ge,yr,4 corr o f (2 Hours) gm gg j 60 minutes to set Low Population M E-4 % g x E-ZD g,,,
circulator seals and Zone Boundary 4,5 ggfy t. con
- 3. 2,
- 400 lbs/ day leakage (30 day) kefors aS until entire primary coolant inventory is Exclusion Area 5.4 x E.3 2.3 x E-2 D&A-2.)-
released Boundary
- 11. 5 0 lbs + 40 / 'M7 (2 Hours) .
10CFR100 Guidelines Low Population . 25 300 -
Zone Boundary *
(Duration of Accident)
- Exclusion Area 25 300 Boundary (2 Hours)
- LPE dose.s revis e d dow~a.d ;- R.< g A es 4 0- A report TOS 493,5- dew 18; 19 74
(
1 q ,
Table 14.13-1 St# MARY OF OFF-SITE DOSES RESULTING FROM POSTULATED ACCIDENTS
/ '
Total Duration Dose (rem Accident Maximum Dose Whole Body Thyroid / Bone Complete Loss of Forced Low Population 3.7 x 10-" 3.6 x 10 2 1.0 x 10-3 Circulation Cooling--DBA No. 1 Zone Boundary Worst PCRV Penetration Exclusion Area 2.5 17.4 4.8 Failure (both closures Boundary of a steam generator penetration)--DBA No. 2
" Maximum Credible Accident" Exclusion Area 0.162 0.088 8.2 x 10-4 (largest potential PORY ~ Boundary leak rate)
Control Rod Withdrawal --
0 0 0 Fuel Handling and Storage -- 0 0 0 Loss of Cooling 2,
a CE LM
I CONCLUSIONS REGARDING CALCULATION OF 0FFSITE -
DOSES FROM HELIUM LEAKAGE CONSERVATIVELY CALCULATED
' LOWER THAN MAxlMLN CREDIBLE ACCIDENT (MCA)
INFSAR(ANDBASEDONWORSEDILUTIONTHANMCA)PREVIOUSLYANA l ,
VERY SMALL FRACTION OF 10CFR100 (0RDERS OF MAGNITUDE LOWER)
O e eV 4
_,_,--___._--.,_...__-c..__, , -..._..r.._.-. . _ . . _ - , _ . . y ,,_ ,_. , , , _ _ _ . , - -- . , _ _ , , , . - _ . _ , . _ . _ , , . , _ _ _ _ - , . . .
n
, , May 7, 1986
-3 Overall, the meeting was very informative. The licensee addressed each of the topics thoroughly, although not always to the staff's complete satisfaction. The staff will review PSC's submittals as a means of resolvino any remaining staff concerns on these topics.
original signed by Charles S. Hinson, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
, ~
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIB'JTLON:
i 500 Ret M11eu: 0
'NRC PDR Local PDR-SSPD Reading
.DCS.
PNoonan CHinson KHeitner Olynch HBerkow OELD EJordan BGrimes ACRS (10)
NRC Participants Q)V WWl /
0 :SSPD DPWRL-B:SSPD DPWRL-8:SSPD DP RL-B:SSPD DPW :SSPD P nan CHinson:ac KHeitner Olynch HBe 5 /86 05/f/86 05/7 /86 05///86 05/') /86
%