ML20214P206
| ML20214P206 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/19/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214P025 | List:
|
| References | |
| FOIA-86-687 NUDOCS 8612040113 | |
| Download: ML20214P206 (13) | |
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- o g UNITED STATES f* n NUCLEAR REGULATORY COMMISSION g ,I wasHINcTow,0. c. 20sss
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SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDING THE CONTAINMENT ISOLATION VALVE TABLE TECHNICAL SPECIFICATION RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458
1.0 INTRODUCTION
In River Bend Technical Specification 3.6.4, Table 3.6.4-1. Item a.1 identifies the reactor core isolation cooling (RCIC) steam supply valve IE51*MOVF076 (MOVF076) as a primary containment automatic isolation valve. By letters dated September 17, 1986 and September 19, 1986, the Gulf States Utilities Company (GSU) requests a one-time change to the Table by adding a footnote, applicable to isolation valve MOVF076, that would exempt the operability requirements for this valve from September 22, 1986 through October 4, 1986.
While the reactor was at 100% power on September 8,1986 the MOVF076 valve was undergoing a functional test of the annunciator when the MOV76 breaker tripped due to a ground in the system. A subsequent attemp't to stroke the valve indicated the valve to be fully open. As required by the River Bend Technical Specification 3.6.4 Action Statements, the outboard isolation valve IE51*M0VF064(MOVF064)wasclosed. Consequently, the RCIC system was rendered inoperable and by Specification 3.7.3, the plant was placed in a 14-day Action Statement tha,t requires the plant to initiate a shutdown on September 22, 1986.
The GSU_ proposed technical specification revision will exempt valve MOVF076 from~ technical specification operability requirements, thereby allowing =
isolation valve F064 to be opened and returning the RCIC system to operable 20 3 861125 GUSTE86-687 PDR
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s',a tus . This request would be a one-time modification to the Technical Specification effective through October 4, 1986, at which time GSU has scheduled a shutdown and will repair the inoperable isolation valve (MOVF076).
2.0 EVALUATION The RCIC steam turbine supply line is an 8-inch line that branches off the main steamline "A". Inside the River Bend drywell the 8-inch line has a inboard isolation valve IE51*MOVF063 (MOVF063)and circumventing this valve is a 3/4 inch bypass line consisting of a manual isolation valve and the motor-operated isolation valve M0VF076. The 3/4-inch bypass line is parallel to the 8-inch inboard isolation valve; its functions during start-up are to equalize i,.he pressure around MOV63 and to warmup the RCIC steamline. The consnon outboard isolation valve (8-inch) is MOVF064. During power operation, the normal RCIC turbine steam supply line valve lineup is MOVF076 closed and inboard and outboard isolation vlaves MOVF063 and MOVF064 open.
The RCIC is a high pressure system primarily designed to maintain 1
sufficient water in the reactor pressure vessel to cool the core and to maintain the reacter in the standby condition in the event .the vessel becomes l 1solated and feedwater is not available. During a LOCA, RCIC (which augments other core injection systems) initiates on low vessel water level and deliverers rated flow to the vessel through a connection in the vessel head.
Even though RCIC is not a part of the ECCS network, the RCIC system is designated as a safety-related system. During limiting conditions of operation (LCO) (i.e.. when HPCS is inoperable), power operation is allowed to continue for a period-of time provided RCIC is operable. ,
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Valve MOVF076 is only used for RCIC turbine steam supply line wann-up before start-up of the system and the ability to close the valve is not required for the RCIC system to perform its intended design function.
Therefore, in the open position MOVF076 will not affect system operation.
The RCIC steam line isolation valves MOVF063 and MOVF064 are normally kept open and should be kept open in order to keep the steam inlet lines warmed-up and to minimize the potential for water hamer upon system startup.
Accordingly, the RCIC steam supply line does not receive the standard containment isolation signals (e.g., high drywell pressure and low reactor water level). The isolation . signals that serve the inboard and outboard isolation valve monitor conditions to reveal a break in the RCIC syrtem. .
The applicant in a letter dated September 19, 1986, also submitted a radiation dose assessment. In the event of a pipe break downstream of the outboard isolation valve (MOVF064), the leakage will be detected by the leakage detection systems. The leakage detection systems are the area temperature l monitoring system and the RCIC steam line flow monitoring system. Either leakage detection system will provide an isolation signal to MOVF063 and MOVF064. Once a signal is received, a total of 20 seconds (a 10 second delay for diesel start plus a 10 second delay for valve closure) delay is assumed for closing of the valve. Full flow is assumed through the 8-inch line for the first 20 seconds. At the end of 20 seconds, the inboard isolation valve F063 is assumed to close, but the outboard isolation valve fails to close (single failure). With the. inoperable valve F076 in the failed open position and the outboard F064 assumed to be failed open, there will be an 1solated M
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1eakage path'(3/4-inch line) from the reactor to the auxiliary building. The e
maximum expected flow through the bypass (F076) is 30 GPM. Feedwater flow could provide make-up for this fluid loss. If feedwater is not available, the High Pressure Core Spray (HPCS) system could be used to provide cooling water to the vessel, thereby preventing uncovering of the reactor fuel. Fuel failure would, therefore, be precluded. The only radioactivity available for release from the break is that which is present in the reactor coolant prior to the break. The bypass flow is assumed to be constant for the first 30 minutes of the transie-; and then is ramped down to zero in the next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as the reactor is brought down to a cold shutdown condition. The total mass released from the RCIC steam lin'e break is estimated to be about 50,000 lbs compared with the 140,000 lbs estimated by the staff for Main Steam Line Break' Accident Analysis. In section 15._6.4 of River Bend SER, we concluded that the doses for Main Steam Line Break are only a small fraction of 10 CFR Part 100 Guideline values. Since the RCIC piping break with unisolated bypass flow through the F076 valve is bounded by the Main Steam Line Break analysis, the expected doses for the RCIC break also will be a small fraction of 10 CFP ""t i
100 guidelines values, and hence within our acceptance criteria.
The staff concurs with GSU's assessment that there is no significant reduction 'in the margin of safety relating to the temporary period (i.e., about 12 days) of having the MOVF076 valve in an inoperable condition. The ,
alternative approach would require reactor shutdown in order to enter th6 drywell and subject plant personnel to exposures from the harsh drywell environment. Therefore, such an action would induce unwarranted and inherent challenges to various plant systems and to the affected plant personnel.
3.0 CONCLUSION
The staff finds the incremental risk of allowing the 3/4-inch inboard isolation valve to be inoperable through October 4, 1986 to be minimal.
Therefore, the staff approves GSU's proposed amendment to the River Bend Technical Specification as indicated in the licensee's September 17, 1986 and September 19, 1986 letters. Specifically, a footnote will be added to the Technical. Specification Table 3.6.4-1 not requiring MOVF076 to be operable through October 4, 1986.
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Enclosure 2 FOB / DBL SALP INPUT PLANT: River Bend Station, Unit 1 LICENSEE: Gulf States Utilities Company DOCKET NO: 50-458 LICENSEE STATUS: OR SER
SUBJECT:
REVISED TECHNICAL SPECIFICATION FOR CONTAINMENT ISOLATION VALVE TABLE PERFORMANCE PARAMETERS:
(1 f) Management Involvement in Assuring Quality
) Approach to Resolution of Technical of Issue From a Safety Standpoint Response to NRC Initiatives
((, )) Staffing (Including Management) q ) Reporting and Analysis of Reportable Events I Training and Qualification Effectiveness L, J) Any Other SALP Functional Area ,
PERFORMANCE NARRATIVE DESCRIPTION OF CATEGORY /
, PARAMETER APPLICANT / LICENSEE'S PERFORMANCE RATING 1 The quality of submittals was 3 adequate 2 The understanding of the issue was 3 adequate 3 Responses were timely ,. 3 4 No basis for assessment N/A 5 No basis for assessment N/A 6&7 No basis for assessment N/A OVERALL APPLICANT / LICENSEE PERFORMANCE RATING 3 e
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MEM0h DUM FOR: Sholly Coordinator DISTRIBitTT6 et File FROM: Walter R. Butler Director PDf4 Reading BWR Project Directorate No. 4 M0'Brien ision of BWR Licensing SStern OGC w/ incoming
SUBJECT:
REQUES R PUBLIC IN BI-WEEKLY FR NOTICE - NOTICE OF CONSIDERAT SSUANCE OF AMENDMENT TO FACILITY OPERATING :
LICENSE AN n D NO SIGNIFICANT HAZARDS CONSIDERATION DETER ION AND OP RTUNITY FOR HEARING Gulf States Utilitie's Company, Docket No. 50-458, River Bend Station, Unit 1 West Feliciana Parish, Louisiana Date of amendment request:
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%;;;-t ' and %,;;t 15,1986.
I'l Description of amendment request: Amend Attech.T.; .t -III er,d Technical Speci-fication S hb. v 0.S. 4 of the River Bend Station Operating License. NPT-47 to i
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l Basis for proposed no significant hazards consideration determination: The Comission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment 1
to an operating license for a facility involves no significhnt hazards considerations if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
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- 2-The licensee has provided an analysis of sig f aza s con iderations in
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its "usud 4,1956 requestfo&A r a license amendm t.
i The licensee has[9 % M lief' concluded, wi.th appropriate bases, that the proposed amendment meets the three :
standards in 10 CFR 50.92 and, therefore, involves no significant hazards considerations.
The Commission has also provided guidance concerning the application of these standards by providing examples of amendments considered likely, and not likely, to involve a significant hazards consideration. These were published in the Federal Register on March 6,1986(51FR7744). The NRC staff has made a preliminary review of the licensee's submittal. A discussion of these examples as they relate to the proposed amendment follows. .
One of the examples of actions involving no si nificant hazards considera ion u-M A R h a n tM M 60. %
yi) is a change thet ccastitutes 4r. Litienel limitatie. , restriction r-ee h c hetec$ais ti m cl1 el 442bu.t J2, to w weud- a y M udigth
=ar: :tringent surveilience requirement. The proposed amVn nt does et involve a nificant increase in the probability or c equence an accident previo ly evaluated; or create the possi lity of a ew or rent kind of accident fr any accident previously valuated; or involv a significant reduction i a margin of saf cause the dherenc to the TDI Diesel Generator Owners Gro DRQR ort requires i pectio that ar more thorough than the manufactur econsnendations those equired y the current Technical Sp ica ons an egulato Guide .108. his adherence to the DRQR rt req res additional ins ctions which re designed to incre the. reliability of the Division I and Divi on II TDI diesel 9enerators. g 0""' "
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1 The proposed amendment to the Technical Specificaticns would not involve a significant increase in the probability or consequences of an accident previously evaluated because the unisolated penetration is limited to a 3/potentially 4" line discharging to a closed system that returns to the containment f vL gesuming
.MOV45, 10CFRa breakPartin the 100 RCIC limitssteamwill line between valves HOV64 and not be exceeded.
conclusion is This fuel failure fluid loss is can not expected because the make-up from waterbased the up Pressure Core Spray system.be provided The only by the Feedwater system or the High release from the break is that which isactivity available for coolant and steam lines prior to the break.present Moreover, in the the reactor break LOCA Recirculation large failure assumes (Double-Ended Line Break) with fuel by the a closed loop in the RCIC system and filtration Standby Gas Treatment and meets the currant accident analysis System for any valve packing leaks limits. ' and 10CFR Part 100 6
Additionall eq'q,sted y mation Lc6'ncer ng WI Q,of-dhe b- ~y6uY J~ ta{WoAignificanVNatardsf -~was7 TopKDm EvhluatichT .
ANtdA4 M The proposed cMnge does not create the possibility of a new or different kind of accident from any accident
. previously evaluated because line breaks such as may be postulated for the RCIC system have bean, considered in the analysis and design basis with many of the previous i analysis bounding all the short-term consequences of the plant for this event. There may,however, be longer term environmental effects which would only impact the overall lifespan of the equipment and would be evaluated if such an event should occur. This change introduces no new mode of operation and no change in the plant design is being made.
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I ouzr erarse verzirzas couraws-aw aeenoerano= poet omes som see er.mancema.miom 4=eA co06eos e8600es Seteset September 19, 1986 RBC- 24405 File No. G9.5 t Mr. Barold R. Denton, Director Office of Nuclear Reactor Rapistion U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear Mr. Denton River Bend Station - Unit 1 Docket No. 50-458 :
As requested by 'your Staff, Gulf States providing below supplemental information to our Utilities (G80) is application for amendment to the River Bend Station specifications, Appendix A to facility Operating Unit 1 License Technical detailed in a letter (J. C. NPF-47 Doddens to R. R. Denton) dated September 17, 1986 (RBG-24405).
The River Bend Station Operations staff performed a functional test on the 1E51*MOVF076 September.8, 1986 when the MOV76 breakerannunciator tripped.at approximately 1415 on investigation determined a ground in the system. The initial following two days, Electrical Maintenance investigated During the the grounding source An attempt was made to dry-out the valve motor. The operator and indicated that the motor was apparently wet.
, then attempted to stroke the valve and indications revealed that the valve was fully opened. An attempt was made to .reclose the valve back to its normal position at which time the McV76 breaker tripped. During the next few days, several other .
made to reclose Mov76 attempts were and to establish valve operability. On September 15, 1986 Electrical Maintenance indicated that it had exhausted all means of fixing 'the grounding problem and proceded to investigata a request for Technical Specification CSU i change.
calculations Betweenand September 15 and September 17 G8U confirmed, by analysis, the safety significance of proposed change. the A hand-drawn schematic that features'the main com are discussed in this request is proivded as Atta ents which your use. nt , 8 for The Remotor Core Isolation Cooling (RCIC) s small in instrument lines ( 3/4") and drain lines (ystem 1") hascontained several rooms that have been analysed for large line* breaks. We have
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. I lines each the examined withRCIC line and have identified seveninstrument 3/4" normally opened valve located in the RCIC room.normally The analyses for opened valves ,
8 team Tunnel and RCIC cubicle areas.these lines include n a Hig Thethe in RCIC main line is common steam tunnel. to the Residual Heat Removal (RNR) line steamBend River condensing Station. mode of operation which is not presently used the RER Equipment Removal cubicle.This RHR line is isolated by two valves in lines each Break analysiswith havenormally been performed opened valves located in the RHR room.The in this Moderate and High Energy Line Crack analysis. area including a Line Breaks.line between MOV64 and Mov45 has been for analysed High The remaining R Energy Assuming MOV45, conclusion a is break in thelimits RCIC will steam notline between valves MOV64 a 10CFR Part ,100 be exceeded. This Fuel failure fluid loss can is not expected because the make-up from the water ~ base Pressure Core Spray system.be The provided by the Feedwater System or the High release from the break is that only which isactivity available for coolantLOCA break and steam lines prior to the break.present Moreover, in the thereactor Double-Ended Recirculation large failure Line by the assume (s a closed loop in the RCIC Break) system with and filtration fuel Ond meetsStandby the current Gas Treatmentaccident System for any valve packing leaks limits. analysis and 10CFR Part 100 Additionally requested by your Staff 1
mation concerning Item 2 of the 'No was supplemental infor-Evaluation": Significant
' Hazards The proposed change does not create the possibility of a new or different kind of accident from any previously accident postulated for the RCIC system have baanevaluated because lin the considered in analysis plant h unding for this event. all the short-term consequences of th i
environmental effects There which may,however, would only impact be longer term the if such an event should occur.overall This lifespan of the equipment and
' no new mode of operation change introduces design is being made. and no change in the plant W
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Therefore, there is no new or different kind of accident from any accident previously evaluated.
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U. C. Deddene Vice President River Band Nuclear Group JCD/JEP/je Attachments e
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ATTACEMENT 1 Assumptions For RCIC Steam Line Break 1
Leakage will be picked up by high temperature leak detection based on 125 gym leakage which provides an isolation signal to F063 and F064. Also high steam flow indication will isolate the RCIC System.
- 2. Once a signal is received to close the isolation valves, a 10 second delay for diesels to start plus a ;
10 second delay for valve closure will be assumed. A total of 20 seconde delay is used. .
3 Assume full flow through the 8" line for 20 seconds.
At the end ofglose assumed'to 20 seconds and the the F064 isolation valve valve fails to F063 is close l
t (single Failure).
second flow from full closure flow to time the 30for gpm the valve bypass willramp flow.y actuallThis the is conse l
4.
Moody critical flow of 2100 lb/sec-ft 2 is used for steam flow rates. flow at 1000 psia will be used to calculate the 5
The bypass flow will be held constant (30 gpm) for the first 30 minutes of the transient and then will be ramp to 0.0 will be in in the coldnext 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at which time the reactor shutdown.
Calculations Mass released for ~
first 20 seconds of = 13,400 lba the transient Mass released for .I next 30 minutes = 7,787 lba of the transient Mass released for next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of = 27,254 1ha the transient .
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T4tal mass from the 48,441 1ha RCIC Break
From River integrated 80 562 lb mass leaving the RPV thBend Station (RBS) FSA on 15.6.4.4 is, based on c.
to the NRC for The radiological analysis for through the steam is line, brea onservative assumptions considered e osteam tline break Basis Accident" Therefore, for RBS as id be and as can meeting be 10CFR100 guidelines is th acceptable valve entified in FSAR Se e " Design case ispiping margin. boundedbreak b< with unisolated ypass flow bseen from the 15.6.4 cacula As previous'r the Main Steam Linethrough st break by thea coF076 the exclusion Table 15.6-7) . area from the i.y stated, the D current FSAR thynsiderable roid dose for conservative Comparing the massesesign Basis Accident is 29 R Ram for the RCIcorrolation to given for each event, a allowable dose.C event which isthebelow dosethewould be approximately 17 30 Rem regulatory i
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. - EMERGENCY LICENSE AUTHORIZATION .
_ CHECK LIST .
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. I. Complete submittal (Section III, Item I)
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final NSHC and
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- a. .ORAB-cr technicah branch input X
- b. Resident or regi'onal personnel input { .
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- 3. *Bes
's Uet .Wb effort" b obtain MWs Muermo P state coments (Section (EP Item III6)- N j b m D u W .d m ae @a.,t%pt ( b eco q
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. Assistant Director concurrence (Section III, Item 7) N g
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k Assistent Director oral authorization to licensee .
(Section III. Item 8)- .
- 6. Telecopy Technical
- Specifications (Section III. Item 8)
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- 7. Forward final two day' . license ' amendment with post notice and FNSHC (Section III. Item'9) (Prepare OLOp 220. Attachment 4) g Y#
,c Project Mana r / .
. Branch Chief - -
ORAB Branch Chief / Tech. Review' Branch Chief * -
ORAS AD,.
/ . Tech. Review Eranch AD*
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