ML20035F622
| ML20035F622 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/19/1993 |
| From: | Oconnor P Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9304220099 | |
| Download: ML20035F622 (41) | |
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4 DocketNo.~-50-416!
April 19, 1993
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i LICENSEEi Entergy 0perations,.Inc.
FACILITY: Grand Gulf Nuclear Station i
SUBJECT:
SUMMARY
OF APRIL.6, 1993, HEETING REGAP. DING REVISIONS TO THE GRAND GULF LICENSE RASIS. LOSS-0F-COOLANT ACCIDENT DOSE METHODOLOGY l
t On April 6, 1993, representatives of-Entergy Operations, Inc. {the licensee),
met with the NRC staff at the NRC headquarters in Rockville, Maryland, to l
discuss Entergy's revisions to the Grand Gulf Nuclear Station license design basis loss-of-coolant accident (LOCA) dose methodology and to identify licensing proposals that may utilize these methodologies in the future. lists the meeting a.ttendees.
~
The licensee provided a detailed discussion of their preparations for future
- applications to eliminate the Main Steamline Isolation Valve (MSIV) Leakage Control System and the feedwater Leakage Coatrol Systems at Grand Gulf.
In addition, the licensee described the initial assumptions, computer code modifications, modeling changes, and dose analysis results associated with the-i t evised LOCA dose methodology at Grand Gulf.
The licensee also offered the NRC staff the opportunity to visit the Grand Gulf site end observe the application of the methodology. contains the visual aids used by the licensee during the meeting.
ORIEINAL SIGNED BY-Paul W. LConnor, Senior Project Manager Project Directorate-IV-1 Division of Rewtor Projects - III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
1.
List of. attendees 2.
Licensee's handout-cc w/o Enclosure 2:
p p 7 g,.7, See next page m a n:a g gy,; _
DISTRIBUTION w/ Enclosure 2:
.iDocket;:{jle2 NRC & Local PDRs
'P. O'Connor PD4-1 Reading E. Merschoff L. Kokajl,
-A. Pelletier J. Hayes J. Lee w/o Enclosure 2
'T. Murley/F. Miraglia J. Partlow J. Roe
- M. Virgilio.
G. Hubbard L. Plisco, EDO P. Hoonan.
OGC E. Jordan
- ACRS (10)
E. Trottier 0FC LA:PD4-h6 PM:PD4-1 k D:PD4 1 h 1
PNoo(anY PO' Connor: pk - U h
NAME-
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OFFICIAL RECORD COPY Document Name: GGKpr693.mts J
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION o
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y April 19, 1993
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Docket.No. 50-416-LICENSEE:
Entergy Operations, Inc.
FACILITY: Grand Gulf Nuclear Station
SUBJECT:
SUMMARY
OF APRIL 6, 1993, MEETING REGARDING REVISIONS TO THE GRAND GULF LICENSE BASIS LOSS-0F-COOLANT ACCIDENT DOSE METHODOLOGY On April 6,1993, representatives of Entergy Operations, Inc. (the licensee),
met with the NRC staff at-the NRC headquarters in Rockville, Maryland, to
. discuss Entergy's revisions to the Grand Gulf Nuclear Station license design basis. loss-of-coolant accident (LOCA) dose methodology and to identify licensing proposals.that may utilize these methodologies in the future.
-Enclosure.I lists the meeting attendees.
The licensee provided a detailed discussion of their preparations for future
-applications to eliminate the Main Steamline. Isolation _ Valve (MSIV) Leakage Control System and..the Feedwater Leakage Control Systems at Grand Gulf.
In-adC tion,.the licensee described the initial assumptions, computer code
. modifications, modeling changes, 'and dose analysis results associated with the revind LOCA-dose rethodology at Grand Gulf. The. licensee also' offered the NRC staff the opportaity to visit the Grand Gulf si_te and observe the
-application of the methodology. contains the visual aids used by the licenta - during the meeting.
Paul W. O'Connor, Sen Manager Project Directorate IV-I Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
- I.
List of attendees
'2.
Licensee's handout cc w/o Enclosure'2:
See next page
?
p 4
a
e--ry-
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w-Tw,
---,.4 r
s Mr. C. Randy Hutchinson Entergy Operations, Inc.
Grand Gulf Nuclear Station CC' Mr. Donald C. Hintz, President and Chief Executive Officer GGNS General Manager Entergy Operations, Inc.
Entergy Operations, Inc.
P. 0.-Box 31995 P. O. Box 756 Jackson, Mississippi 39286-1995 Port Gibson, Mississippi 39150 Robert B. McGehee, Esquire The Honorable William J. Guste, Jr.
Wise, Carter, Child & Caraway Attorney General P. O. Box 651 Department of Justice Jackson, Hississippi 39205 State of Louisiana P. O. Box 94005
. Nicholas S. Reynolds, Esquire Baton Rouge, Louisiana 70804-9005 Winston & Strawn 1400 L. Street, N.W. - 12th Floor Alton B. Cobb, M.D.
' Washington, D.C.
20005-3502 State Health Officer State Board of Health Mr. Sam Mabry, Director P. O. Box 1700 Division of Solid Waste Management Jackson, Mississippi 39205 Mississippi Department of Natural Resources Office of the Governor.
P. O. Box 10385 State of Mississippi Jackson, Mississippi 39209 Jackson, Mississippi 39201 President, Mike Morre, Attorney General Claiborne County Board of Supervisors Frank Spencer, Asst. Attorney General Port Gibson, Mississippi 39150 State of Mississippi Post Office Box 22947 Regional Administrator, Region II Jackson, Mississippi 39225 U.S. Nuclear Regulatory Commission 101 Marietta St., Suite 2900 Mr. John P. McGaha Atlanta, Georgia 30323 Vice President, Operations Support Entergy Operations, Inc.
Mr. Michael J. Meisner P.O. Box 31995 1
Director, Nuclear Licensing Jackson, Mississippi 39286-1995 Entergy Operations, Inc.
P. O. Box 756 Mr. S. S. Chan Port Gibson, Mississippi 39150 Project Engineer Bechtel Power, Corporation Mr. K. G. Hess P.O. Box 2166 t
Bechtel Power Corporation Houston, Texas 77252-2166 P. O. Box 2166 Houston, Texas 77252-2166 i
Mr. Rudolph H. Bernhard r
Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 399
' Port Gibson, Mississippi 39150 l
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ENCLO'SURE 1 3
April 6.1993 LOCA Dose Meetina Grand Gulf Nuclear Station
-l List of Attendees l
Paul O'Connor NRR/PDIV-1 Allison Pelletier NRR/PDI-l.
John J. Hayes NRR/PRPB-
. ]
Jay Lee-NRR/PRPB Ed Trottier NRR/PDI '.i l
Mike Meisner E01/GGNS Dan Pace E01/GGNS Mike Withrow E01/GGNS--
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Bryan Ford E01/GGNS
. Lawrence E. Kokajko NRR/SPLB Marvin Morris Omega Technical Services i
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GRAND GULF NUCLEAR STATION
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PRESENTATION TO THE NRC/NRR STAFF APRIL 6,1993 REVISIONS TO THE GRAND GULF LICENSE BASIS-LOCA DOSE METHODOLOGY elNTRODUCTION AND BACKGROUND MIKE MEISNER I
e PLANT DESIGN OVERVIEW MIKE WITHROW s CURRENT LICENSE BASIS MIKE WITHROW-e
SUMMARY
OF CHANGES MIKE WITHROW.
e DOSE ANALYSIS RESULTS MIKE WITHROW eFUTURE APPLICATIONS BRYAN FORD GRAND GULF NUCLEAR STATION N
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P BACKGROUND eGRAND GULF DESIGN / LICENSE BASIS FOR CALCULATING ACCIDENT DOSE CONSEQUENCES 4
HAS NOT CHANGED SIGNIFICANTLY SINCE PRELICENSING DAYS.
^
eCALCULATIONAL TOOLS AND UNDERSTANDING OF PHENOMENA HAVE ADVANCED CONSIDERABLY.
l e COMMERCIAL NUCLEAR ENVIRONMENTHAS
. UNDERGONE MAJOR CHANGES.
1 e ENTERGY OPERATIONS AND GRAND GULF MANAGEMENT PHILOSOPHY HAS UNDERGONE SIMILAR CHANGES.
i GRAND GULF NUCLEAR STATION
GRAND GULF
~
OPERATING i
PHILOSOPHY e LONG TERM SUCCESS REQUIRES EQUAL EMPHASIS ON SAFETY, COST AND PLANT AVAILABILITY.
e STRONG SAFETY PERFORMANCE REQUIRES, AMONG OTHER THINGS, THE ANALYTIC CAPABILITY TO DETERMINE SAFE THRESH 0LDS WITH APPROPRIATE l
CONSERVATISM.
eGOOD COST PERFORMANCE REQUIRES US l
TO MAINTAIN OR RESTORE THE BALANCE BETWEEN COST AND SAFETY BENEFIT.
GRAND GULF NUCLEAR STATION i
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PURPOSE OF LICENSE l
BASIS CHANGE e EXCESS MARGIN IN ORIGINAL DOSE i
l CALCULATION METHODOLOGY PREVENTED ACCURATE ASSESSMENT OF SAFETY SIGNIFICANCE AND CONTINUING COST i
BURDENS WITHOUT COMMENSURATE SAFETY BENEFIT.
oSTANDBY GAS TREATMENT SYSTEM BYPASS LEAKAGE EVENT DURING RFOS HIGHLIGHTED THE NEED TO MOVE FORWARD ON REVISING THE DOSE CALCULATION DESIGN AND LICENSE BASIS.
GRAND GULF NUCLEAR STATION
~
NATURE OF LICENSE BASIS CHANGE e MAJOR REVISION OF THE METHODOLOGY, ASSUMPTIONS AND' INPUT DATA USED IN i
DETERMINING THE DBA LOCA DOSE CONSEQUENCES.
eIMPl EMENTED UNDER 10CFR50.59.
eINCLUDED IN THE NEXT UPDATE TO THE GRAND GULF FSAR.
- WILL FORM THE. BASIS FOR FUTURE REVISIONS TO THE GRAND GULF LICENSE BASIS AND RESOLUTION OF SAFETY ISSUES.
GRAND GULF NUCLEAR STATION
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PLANT DESIGN OVERVIEW MARK 111 - Pressure Suppression Containment Directs any primary system leakage through the drywell vent system into the suppression pool where the steam is condensed.
Design prevents any significant pressurization of the containment.
Design features to control the release of fission products Drywell Seismic category I structure designed to maintain its integrity during and following any postulated LOCA including the worst single failure in conjunction with a loss of offsite power.
Designed to limit the energy and fission product transfer to the containment by directing the flow from postulated pipe ruptures through the suppression pool..
GRAND GULF NUCLEAR STATION
PLANT DESIGN OVERVIEW The drywell design pressure is 30 psig and the design temperature is 3300F.
During a DBA-LOCA, primary system leakage is directed through the horizontal vent system into the suppression pool (135 horizontal vents arranged in.three rows).
The allowable drywell bypass leakage is based on the amount of steam bypass which can be allowed without exceeding the containment design pressure of 15 psig.
The allowable leakage is expressed in terms of the parameter ANK where A 2
is the flow area of the leakage path (ft ) and K is the geometric and friction loss coefficient.
The A/VK value is 0.90 ft.
2 GRAND GULF NUCLEAR STATION
PLANT DESIGN OVERVIEW Containment Designed to limit the leakage from the containment free volume to 0.35 %/ day.
Containment integrity is demonstrated per Technical Specification 3.6.1.2 to be less than 0.75 of the allowable leakage La (0.35 %/ day) when tested at the peak accident pressure (Pa) of 11.5 psig.
Inclusion of the MSIV leakage (100 scfh, total of four steam lines) results in an equivalent containment leakage of 0.437 %/ day.
The suppression pool is designed as a passive means of removing blowdown energy resulting from a postulated design basis LOCA.
The suppression pool is effective in removing radiciodine released during
.the design basis LOCA.
GRAND GULF NUCLEAR STATION 4
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PLANTDESIGN OVERVIEW Containment Sprays Sprays are designed to limit containment pressurization resulting from steam leakage which bypasses the suppression pool and are also effective in removing radiciodine which could be present in the bypass steam.
L Separate and redundant spray trains are provided which are automatically p
initiated upon high containment and drywell pressure.
a Hydrogen Mixing System 0
li The system is designed to purge the hydrogen produced within the drywell into the containment through the suppression pool at a design flowrate of 500 'scfm/ train.
Pressurizes the drywell until the first row of horizontal vents is uncovered allowing flow of the drywell atmosphere through the suppression pool and into the containment.
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r PLANT DESIGN OVERVIEW MSlV Leakage Control System Designed to control and minimize the main steamline leakage which could bypass the secondary containment.
Directs any leakage past the MSIVs into an aux. building area served by the SGTS.
Manually' actuated at approximately 20 minutes post-LOCA.
GRAND GULF NUCLEAR STATION
PLANT DESIGN OVERVIEW STANDBY GAS TREATMENT SYSTEM Designed in accordance with the requirements provided in 10 CFR Paris 50 and 100.
Controls unfiltered leakage and minimizes post accident offsite doses to the public.
Design bases of the SGTS and. secondary containment are described as follows:
a.
Limits the thyroid and the whole body doses to less than the limits of 10 CFR 100 and 10 CFR 50 General Design Criterion 19.
b.
Develops and maintains a 1/4-inch w.g. negative pressure in the boundary region within 120 seconds after actuation.
c.
.. Designed as a Seismic. Category.1 system.
GRAND GULt: NUCLEAR STATION
0-4 PLANT DESIGN OVERVIEW STANDBY GAS TREATMENT SYSTEM d.-
Provides a. dilution.and hold-up volume for fission products which may leak from the primary containment following a postulated accident.
e.
Designed to permit periodic inspection and testing of principal system components.
f.
Two 100 % redundant SGTS trains are provided..
g.
Designed to be ' OPERABLE with either on-site or off-site power.
available.
h.
Initiated by high drywell. pressure,-low RPV water level, high radiation in the fuel handling area, or: manual signals.
i.
Meets the' requirements of RG 1.52 for a 99% removal efficiency for elemental and particulate iodine.
GRAND GULF NUCLEAR STATION
PLANT DESIGN OVERVIEW CONTROL ROOM STANDBY FRESH AIR SYSTEM Automatically isolated on high radiation, low reactor water level, or high drywell pressure signals.
Following isolation, the exhaust and makeup valves close and the system is operated in a recirculation mode.
Two full capacity. redundant trains each containing a 4,000 scfm fan, charcoal filters, and the associated ductwork, dampers, and controls are provided.
Fresh air units draw 4,000 scfm-of the 33,000 scfm of air recirculated by the control room air conditioning units.
Recirculation air. passes through'a demister, heating coils, pre-filter, a HEPA filter, an activated charcoal filter, and a second HEPA filter.
GRAND GULF NUCLEAR STATION
PLANT DESIGN OVERVIEW CONTROL ROOM STANDBY FRESH AIR SYSTEM An isolation signal initiates both fresh air systems.
One of the systems may then be manually placed in the standby mode.
Control room filters are designed to meet the requirements of Regulatory Guide 1.52 with an iodine adsorption efficiency of 95%.
Drywell integrity verified during periodic testing to demonstrate that the bypass leakage is less than or equal to 10 % of the acceptable A/VK bypass leakage 2
design value of 0.90 ft.
GRAND GULF NUCLEAR STATION
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REGULATORY REQUIREMENTS I
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= ASSESSMENT OF RISK TO PUBLIC HEALTH AND f
SAFETY REQUIRED BY SECTION 50.34 OF 10CFR PART 50.
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= DOSE EVALUATION ' ACTORS ARE PROVIDED IN
'10CFR100, PART 100.11.
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= 10CFR50, APPENDIX A, GENERAL DESIGN i
l CRITERlON 19 PROVIDES REQUIREMENTS ON DOSES TO CONTROL ROOM PERSONNEL.
10CFR PART 50, APPENDIX A, GENERAL DESIGN j
CRITERIA 16,41,42, AND 43 PROVIDE REQUIREMENTS CONCERNING CONTAINMENT l
AND ATMOSPHERIC CLEANUP SYSTEMS.
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REGULATORY GUIDANCE 1
e REGULATORY GUIDE 1.3, REVISION 2 PROVIDES ACCEPTABLE i
ASSUMPTIONS AND METHODS FOR DEMONSTRATING i
COMPLIANCE WITH THE REGULATORY REQUIREMENTS OF 10CFR PARTS 50'AND 100.
a BRANCH TECHNICAL POSITION CSB 6-3 ESTABLISHES ADDITIONAL REQUIREMENTS RELATIVE TO THE DETERMINATION OF LEAKAGE WHICH BYPASSES THE SECONDARY CONTAINMENT.
e STANDARD REVIEW PLAN (SRP) 6.5.5 ALLOWS CREDIT FOR THE F
PRESSURE SUPPRESSION POOL AS A FISSION PRODUCT j
CLEANUP SYSTEM. THIS GUIDANCE REPLACES THAT PROVIDED l
lN REGULATORY GUIDE 1.3 POSITION C.1.f.
e REGULATORY GUIDE 1.52 PRESENTS METHODS ACCEPTABLE TO THE NRC STAFF FOR IMPLEMENTING THE COMMISSION'S REGULATIONS IN APPENDIX A TO 10CFR50 WITH REGARDTO ENGINEERED-SAFETY-FEATURE (ESF) ATMOSPHERIC CLEANUP SYSTEMS.
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METHODOLOGY CHANGES e OFFSITE DOSES WERE CALCULATED USING THE GGNS " TRANSACT" COMPUTER CODE WHICH IS DERIVED FROM THE " TACT 5" COMPUTER CODE.
r
-CODE VERIFIED IN ACCORDANCE WITH GGNS o
PROCEDURES.
-ORIGINAL ANALYSIS USED THE "LOCADOSE" CODE WHICH WAS BASED ON THE " TACT lil" COMPUTER
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CODE.
e OFFSITE X/Q VALUES WERE CALCULATED WITH THE
" PAVAN" COMPUTER CODE.
-CODE.WAS MODIFIED BY GGNS TO EXECUTE ON A L
PERSONAL COMPUTER.
-VERIFIED IN ACCORDANCE WITH GGNS I
PROCEDURES.
GRAND GULF NUCLEAR STATION
i ASSUMPTION CHANGES e DOSE CONVERSION FACTORS - CHANGED FROM ICRP 2 AND REG. GUIDE 1.109 TO ICRP 30 BASED ON THE PROPOSED 10CFR20.
e RADIONUCLIDE REMOVAL BY SUPPRESSION POOL SCRUB'BING - ADDED FOR ELEMENTAL AND PARTICULATE LODINE (DF=10) BASED ON SRP SECTION 6.5.5.
e ADDED AN ALLOWANCE FOR UNFILTERED SECONDARY CONTAINMENT LEAKAGE (50 SCFM SGTS BYPASS FLOW).
GRAND GULF NUCLEAR STATION
~
INPUT PARAMETER OHANGES e METEOROLOGICAL DATA - REVISED OFFSITE X/Q VALUES BASED ON UPDATED SITE DATA (1988 THROUGH 1991).
o CONTROL ROOM INLEAKAGE RATE -
REDUCED FROM 760 SCFM TO 590 SCFM (+10 SCFM FOR DOOR OPENINGS) BASED ON THE CANCELLATION OF GGNS UNIT 2.
e CONTROL ROOM VOLUME - INCREASED FROM 165,000 CU. FT. TO 253,000 CU. FT. DUE TO VARIOUS DESIGN AND CALCULATION UPDATES.
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2 DOSE ANALYSIS
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RESULTS 5% Probability Level X/Q Values (sec/cu. m)
OLD X/Q VALUES TmEI PEPJOO5 OftS) 0-2 2-8 8 - 24 24 - 96 96 - 720 SITE (696m) 1.08 - 03 LPZ (3219m) 3.26 - 04 2.91 - 04 4.22 - 05 3.33- 05 2.32 - 05 NEWXIQ VALUES Tm4? PUnto(HRS) 0-2 2-8 8 - 24 24 - 96 96 -720 SITE (696m) 1.26 - 03 LPZ (3219m) 2.84 - 04 1.56 - 04 1.16 - 04 6.06 - 05 2.39-05 TIME PERIOOS -
0 - 2 min 2 min - 8 hrs 8 - 24 hrs 24 - 96 hrs 96 - 720 hrs CONTROL ROOM 1.34 - 03 3.29 - 04 1.96 - 04 1.08 - 04 6.98 - 06 GRAND GULF NUCLEAR STATION
DOSE ANALYSIS
~
RESULTS Offsite (LPZ, EAB) and Control Room (CR) Doses (Rem?
THYROID WHOLE BODY LPZ EAD CR LPZ EAB CR 10 CFR LIMITS 300 300 30 10 CFR LNITS 25 25 5
ORIGINAL ANALYSIS 73.91 103.4 29.54 ORIGINAL ANALYSIS 15.81 12.42 0.67 REVISED ANALYSIS 61.66 18.84 12.20 REVISEDANALYSIS 0.634 1.372 0.027 1
GRAND GULF NUCLEAR STATION
i FUTURE APPL CATIONS f
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1 GRAND GULF NUCLEAR STATION
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POTENTIAL FUTURE
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APPLICATIONS o EMERGENCY PLANNING DOSE ESTIMATION.
e ELIMINATION OF MSIV LEAKAGE CONTROL SYSTEM.
- e. ELIMINATION OF FW LEAKAGE CONTROL SYSTEM.
e SECONDARY CONTAINMENT GRAND GULF NUCLEAR STATION
OBJECTIVES OF MSIV LEAKAGE CONTROL ELIMINATION / MODIFICATION PROGRAM elMPROVE PERFORMANCE OF THE MSIV LEAKAGE CONTROL SYSTEM.
e LOWER DOSE EXPOSURE TO MAINTENANCE PERSONNEL (ALARA).
e RESOLVE PROBLEMS WITH OBTAINING MSIV LEAKAGE CONTROL SYSTEM REPLACEMENT COMPONENTS.
e EXTEND EFFECTIVE MSIV SERVICE LIFE BY-INCREASING ALLOWABLE LEAKAGES.
e MINIMlZE CRITICAL PATH OUTAGE EXTENSIONS DUE TO:
-REPAIR OF THE MSIVs: AND MSIV LEAKAGE CONTROL. SYSTEM.
-PRECLUDE REPEATED LOCAL LEAK RATE TESTING.
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