ML20062L132

Forwards Info Re post-TMI Requirements Listed in NUREG-0737

ML20062L132
Person / Time
Site:	FitzPatrick
Issue date:	01/08/1981
From:	Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.A.1.3, TASK-1.A.1.4, TASK-1.C.1, TASK-2.B.1, TASK-2.B.2, TASK-2.B.3, TASK-2.B.4, TASK-2.E.4.2, TASK-2.K.3.27, TASK-TM JPN-81-5, NUDOCS 8101160441
Download: ML20062L132 (43)

Text

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?OWER AUTHORITY OF THE STATE OF NEW YORK to CoLUMeus CIRCLE NEW YORK. N. Y. loo 19 (212) 397 620c.

Gron,G,Ey. sERRY OPE R ATING OFFICER TRUSTEES JOHN W. BOSTON JOHN S.QYSON RE soE a atCTOR OF POWER CPER ATIONS GEORGE L :NGALLS JCSEPH R. SCHMIEDER vics cnasRMAN ES DE

&C itF RICHARD M. FLYNN LERQY W. SINCLA RQ8ERTl. MILLO N ZI rREoERiCx R. cLARx January 8, 1981 j,cl;;';"^"ci^'

JPN-81-5 ysou,,,.

,,,y Director of Nuclear Reactor Regulation

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U. S. Nuclear Regulatory Commission Washington, D. C.

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Attention:

Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 2 Division of Licensing

Subject:

James A.

FitzPatrick Nuclear Power Plant Docket No. 50-333 Post-TMI Requirements j

4

Reference:

Letter, D. G. Eisenhut (NRC) to all Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits dated October 31, 1980

Dear Sir:

contains a submittal of information and/or discussion of the following items as listed in NUREG 0737:

1.A.l.l.3 Shift Technical Advisor 1.A.l.1.4 Shift Technical Advisor,Long-Term Program 1.C.1 Short Term Accident and Procedures Review II.B.1 Reactor Coolant System Vents II.B.2 Plant Shielding II.B.3 Post Accident Sampling Capability II.B.4 Training for Mitigating Core Damage II.E.4.2.5 Containment Isolation Dependability-Containment Setpoint Pressure II.E.4.2.6 Containment Isolation Dependability-Containment Purge Valves II.F.2 Instrumentation for Inadequate Core Cooling II.K.3.3 Safety Relies Valve Challenges Since April 1, 1980 II.K.3.13 Separation of HPCI and RCIC System Initiation Levels II.K.3.15 Modify Break-Detection Logic to Prevent 00/

Spurious Isolation of HPCI and RCIC i

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THIS DOCUMENT CONTAlHS m 1160 94/

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POOR QUALITY PAGES

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' II.K.3.17 Report on Outager of Emergency Core-Cooling Systems II.K.3.21 Restart of Core Spray and LPCI Systems II.K.3.22 Automatic Switchover of RCIC System Suction II.K.3.27 Provide Common Reference Level for Vessel Level Instrumentation II.K.3.44 Adequate Core Cooling for Transients with a Single Failure 4

II.K.3.45 Evaluation of Depressurization with other than Automatic Depressurization (ADS)

III.A.2.1 Emergency Preparedness i

III.A.2.2 Meteorological Measuring System III.D.3.3 Improved Inplant Iodine Instrumentation Under Accident Conditions III.D.3.4 Control Room Habitability Submittal of information required by items II.E.4.2.6, III.A.2.1 and III.D.3.4 will be made by the Power Authority in separate transmittals.

Information required by items II.K.3.13, II.K.3.21, II.K.3.27, II.K.3.44 and II.K.3.45 will be submitted through the GE/BWR_ Owners' Group.

The information contained above and 4

in Enclosure 1 was discussed with NRC staff in a telephone con-versation on January 2, 1981.

Should you require further information, please do not hesitate to contact us.

Very truly yours, h.h $

V' J.

P. Bayne Senior Vice President Nuclear Generation r

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ENCLOSURE 1 POST-TMI REQUIREMENT 3 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A.

FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 JANUARY 8,

1981

l.A.l.l.3 SHIFT TECHNICAL ADVISORS The James A. FitzPatrick plant staff contains six (6) Shift Tech-nical Advisors (STAS) all of whom possess a bachelor's degree in an engineering or related scientific discipline. The following is a description of the training program completed for these personnel on December 18, 1980. Additional detail regarding course content is retained by the facility.

Phase I Plant System and Operations This phase consists of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of instruction in the design, in-teracticn and responses of the major plant systems. Classroom instrr.ction was followed by a minimum of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of programmed and supervised oa-the-job training leading to familiarization with physical systems and their operation. Emphasis was placed on those systems which play an important role during analyzed accidents and transients. This phase was taught the level of knowledge required of a reactor operator.

Phase II -

Nuclear Engineering This phase was taught at the Reactor Analyst level and censisted of, 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of instruction in reactor physics, reacter control, thermal-hydraulics, core design, core management and the process computer. The objective was to provide the STA'n with detailed understanding of reactor operation and surveillance.

Phase III - Abnormal Event Analysis This phase consisted of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom instruction in the response of the reactor and reactor plant to analyzed accidents and tran-sients with special emphasis on reactor and core response.

Incl'uded were the thermal and pressure safety limit ef fects of selected failures atid common-mode failures.

Instrument response and backup instrumentation were highlighted.

Ph;se IV

- Simulator Training

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Each STA received 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> of training at the TVA Brown's Ferry Simulator Facility. The program culminated in the " hot" license certification of NUREG-0094, Appendix F.

Four hours of each day were devoted to class,

room review and discussion, and four hours devoted to simulator operationp.

The STA's concentrated on normal evolutions during the first part of thef course, and on transient and accident behavior during the latter part.

l Phase V

- Leadership and Management This phase consisted of 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of management seminars designed j

to satis fy all of the recommendations in NUREG-0737, [ Appendix C, paragraph 6. 3.

The 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> is distributed as follows:

1.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of interactive instructions'in_ situation problem analysfs, decision analysis and potential problem analysis applied to the

,1 Control Room environment.

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2.

32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> of interactive instruction in leadership, motivation, effective communication and human relations.

Included were introductions to several personal management styles.

3.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of interactive instruction in the causes, impact, reactions and management of stress.

Rcqualification of STA's will be conducted, in lieu of more definitive guidance, in a manner similar to that of licensed personnel. They will attend classes on a quarterly basis, and will be retrained in the areas of interest to their specific job function.

Encisure 1, page 2

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P I.A.l.l.4 S!!IFT TECHNICAL ADVISOR LCNC.-TERM PROGRAM

• s sd A long-term STA program will not differ substantially from the present program in the areas of selectim' ciiteria, qualification and training.

INPO guidelines, or other definitive, approved guidance, will 4i+

be used as the basis for the long-term program.

b; Phase-out of the STA program will depend on the upgrading of Shif t Supervisors and senior nuclear, cperators to the educational requirements eventually to be adopted and issued by the Nuclear Regulatory Commission.

N It is anticipated that corporate policy will be developed regarding the STA long-term program on or about April 1, 1581.

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1.C.1 SHORT-TERM ACCIDENT AND PROCEDURE REVIEW The Authority has been participating in the GE/BWR Owners Group and adopts the positions taken by this Owners Group regarding 1.C l.2 and 1.C.l.3, " Guidance for the Evaluation of Inadequate Core Cooling" and

" Transients and Accidents", respectively. As noted in NUREG 0737, the analysis and guidelines have been submitted by the GE/BWR Owners Group and have been reviewed and approved for trial implementation on six (6)

BWR plants.

If after this trial implementation, the guidelines are found to be acceptable to the NRC staff, the Authority will revise its procedures accordingly., page 4 l

11.B.1 REACTOR COOLANT SYSTDI VENTS The Auttarity addressed the Nuclear Regulatory Commission staff concerns regarding Item ll.B.1, " Reactor Coolant System Vents", in previous correspondence (letter from P. Early (Power Authority) to D. G. Eisenhut (NRC) dated October 22, 1979).

In view of what has been stated in the correspondence, the Authority believes that adequate reactor coolant system venting is provided by the existing pl' ant design.

2ncisoure 1, page 5

9 ll.B.2 PLANT SHIEIDING The Authority through its architect-engineer has performed a shielding-design review in accordance with the requirements of NUREG 0578. A re-evaluation of this shielding-design review is in progress in order to assess the impact of the new requirements contained in NUREG 0737. This reevaluation is expected to be complete by April, 1981.

Any recommended plant modifications identified as a result will be reviewed by the Author-ity's staff and implemented as necessa'ry., page 6

ll.B.3 POST-ACCIDENT SAMPLING CAPAP.ILITY The Authority has purchased and is at present performing the engi-neering review of a post-accident sampling system. The system is being designed to meet the NUREG 0737 requirements with the following exception:

Flow restrictions will not be used to limit coolant loss from ruptured sample lines from the RHR pump discharge samples as recommended in 11.B.3.ll(a).

These are low pressure sample lines and rather than using flow restrictions, we plan to use fully qualified solenoid valves to provide remote isolation capability., page 7 i

i

II.D.4 TRAINIIE FOR MITIGATIl0 CORE DAMAGE Facility procedures have been revised to include this program in the training and retraining of licensed personnel. Development of specific program content awaits data and guidance from the NSSS vendor and the BWR

)wners' Group. The Authority intends to initiate the program for appropriate plant personnel no later than April 1,1981.

., page 8 o

I

e ll.E.4.2.5 CONTAI!EENT ISOLATION DEPENDABILITY CONTAI?NENT SETPOINT PRESSURE The James A. FitzPatrick Nuclear Power P1)it containment setpoint pres-sure, that initiates containment isolation, is currently required to be less than or equal to 2.7 psig. Under normal operating conditions the containment drywell is slightly pressurized with N2 in order to provide a drywell to suppression pool differential pressure of greater than 1.7 psid.

Thus normal containment pressure is approximately 1.8 to 1.9 psig.

The isolation setpoint of less than or equal to 2.7 psig is considered by the Authority to be the minimum practicable pressure and is consistent with the margin of 1 psi recommended by NUREG 0737.

Therefore, no change in setpoint is necessary.

, page 9 l

l

II.E.4.2.6 CONTAINMENT ISOLATION DEPENDABILITY -

CONTAINMENT PURGE VALVES This informaticn is being submitted under separate cover in response to the letter from T.A.

Tppolito (NRC) to G.T.

Berry (Power Authority) dated July 18, 1980., page 10

II.F.2 INSTRUMENTATION FOR INADEQUATE CORE COOLING The Power Authority endorses the position of the GE/BWR Owners' Group on this topic as contained in Section 3.5.2.3 of General Electric report NEDO 24708 and in the letter from D. B. Waters (BWR Owners' Group) to D. G.

Eisenhut (NRC) dated October 8, 1980., page 11

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II.K.3.3 SAFETY RELIEF VALVES CHALLENGES SINCE APRIL 1, 1980 Since April 1, 1980, the only relief valve challenges were on October 13, when valves G, H,

L, and J lifted due to high pressure resulting from a reactor scram.

(G & H 10 times in auto: J 5 times in auto, L 1 time in auto; G S H 1 time in manual). Approximately 50%

of the valves (topworks) were steam tested at Wyle Labs as required by Technical Specificatiens and all relief valves were tested at about 150 psig as part of the post-refueling outage startup-test program., page 12

ll.K.3.13 SEPARATION OF HPCI AND RCIC SYSTEM INITATICN LEVELS As a member of the GE/BWR Owners Group, the Power Authority endorses the position taken by this group regarding this issue. The potential for reducing thermal cycles by separating the RCIC and HPCI initiation setpoint was examined. The results of the analysis indicated that no significant reduction in thermal cycles is achievable by separating the setpoints. An analysis was also performed which evaluates the proposed logic change for the RCIC System automatic reset / restart. This evaluation concludes that such a change contributes to improved system reliability and that it could be accomplished without adverse affect on system function and plant safety.

l The Authority has initiated action and expects to meet the July 1, 1981 l

implementation date.

Information concerning the separation of HPCI/RCIC initiation setpoints was previously submitted by the BWR Owners' Group in a letter from R.

Buchholc (GE) to D. Eisenhut (NRC) dated October 2, 1980.

It is the Authority's understanding that further information on automatic reset / restart of the RCIC System will be submitted to the NRC from the BWR/ Owner's Group by January 1, 1981.

!, page 13

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11.K.3.15 MODIFY BREAK-DETECTION LOGIC TO PREVENT SPURIOUS ISOLATION CF HPCI AND RCIC The Authority completed a plant modification (No. F1-74-11) in March, 1975 which consisted of installation of a combination of a fluid snubber and a larger ranged instrument on the HPCI and RCIC high steam flow sensors.

Operating history Oyath testing and actual initiations of the systems) from 1975 to the present has demonstrated the effectiveness of this modification in eliminating spurious isolations of these systems., page 14

f

• 6 11 K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE-COOLING SYSTEMS Attached is the descriptive report on emergency core-cooling system or component outages for the years 1976-1980. The outages are keyed to applicable FitzPatrick Licensee Event Reports (LER) and for the most part do not reflect maintenance performed when the systems or components were not required by Technical Specifications.

As there are no requirements or up until now a reason for recording actual outage durations, these times are the best estimates where available.

Additionally, the completion dates and duration times are affected by changing Technical Specification conditions such as entry into refuel outages or system lineups which can allow for lengthy or non-critical repair times.

A review of the report, in order to propose changes to improve t he availability of ECCS equipment results in certain commencs:

(a) The overall ECCS availability when required by Technical Specifications appears to be very good with the exception of certain recurrent problems - such as:

(1)

Diesel Generators - which has resulted, during the past few years, in the implementation of a comprehensive preventive maintenance program.

(2) Target Rock Relief Va1.ves (well documented) - which resulted in the total replacement to.a new two-stage model.

(3)

Instrument Drifts - which has resulted in modifications to surveillance schedules and engineering for new style instrumentation.

(4) Motor Operated Valve Limitorque Problems - which has I

resulted in the implementation of a comprehensive Limitorque training program for maintenance electricians, and preventive maintenance program.

(5)

LPCI Battery Inverter - further design and engineering for a better more reliable system needs to be performed.

(b) Same unavailability was due to extensive hanger work performed as part of the Bulletin 79-02,07,11 effo rt. Certain systems were administrative 1y declared inoperative when working a seismic hanger - system was still available for service.

As recurring problems are identified, they are analyzed and solutions identified. Preventive maintenance and training programs provide key elements in retaining a high ECCS availability. Additionally, the present philosoply and guidelines being adopted by the NRC towards the ASME Code Section XI Pump and Valve Test Program are positive steps toward reducing ECCS unavailability due to excessive surveillance tests., page 15

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ll.K.3 17 REPORf ON OUfAGES OF LNERGENCY CORE-C00E ING SYSif MS LICENSEE REPOPT t

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1.1.H/uks. liAll; tillM.H i t'110N CAllSt.

OIMillCflVE ACitoN jo 33 pygq:g innel a ltn.

. -...... 05 760816 Due to relief valve operation and RCit.

Safety relief valve li f t iens.

Pumped down torus to radwaste 760323 turbine exhaust after a aunual scram, during initial phases of LPRtt outage.

LPRH

. torus' level above auximum allowable.

Outage 1

16 760321 While performing ST-24C RCIL flow rate Manual isolation exhaust valve was Renoved and replaced rupture discs.

Best test, the turbine was started, closed.

Estimate l

enhaust pressure, increased and made 760327 j

rupture discs blew out.

76-07 760330 While investigating ground on 02-RV-71 K Fuses were removed to clear ground.

Repaired ground and reinstalled fuses.

760728 the control power fuses were pulled V.stve conid still perform relief disabling the electrical operation of function.

the safety valve.

76-10 760403 During surveillance testing (ST), the instrument drift.

Recalibrated Ismediately.

760403 following switches were found to be a auximum of 0.05 psig. higher than i

Technical Specification requirements of < 20 psig:

10-Il9A, C, 1010, 1008, C, D, 5-17A, B t. D.

t' 76-12 2

'760405 During ST, pressure switch 2-3-55D Nechanical stop on bourdontube.

Adjusted mechanical stop so micro 760406 would not function.

switch contact could be made.

I 76-30 760%I4 During ST-4E, the HPCl gland seal Hormal wear.

Replaced gasket.

760414 6 hrs.

condenser cooling slide gasket blew

.out.

l 76-3) 760420 During ST-4E, the motor for 23-M9V-57 Short in limitorque.

Replaced m tor. In the interim, icpt.

760514 (HPCI torus suction valve) Irurned up valvo open to perfurm function.

when given an open sIgnaI,

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REPORT ON OUTAGES O EMERGENCY CORE-COOLING SYSIENS LIEENSEE REPORI r

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After several attempts, valve closed 760424 3

exercise, 13-HOV-15 failed to close.

within the specified time frame and During the test, the outboard event could not 6e duplicated.

Isolation valve 13-HOV-16 did operate.

I l 16-45.

760713 During ST-2G, 10-50V-2638, RllR process incorrect test signals Corrected procedure.

760713 sample line failed to isolate when signaled to do so.

76-35 760504 RHR pump "C" breaker was racked out Operator error.

Surveillance testing was inanediately 760504 for maintenance thus disabling a initiated.

i port ion of the RHR system, i

76-21 760519 During ST-9B, when Initiating "B" and failed speed toch relays.

Replaced.

760519

.I hrs.

"D" emergency diesels, the "D" diesel would not start.

i 76-24 760611 Annunciator alarmed on HPCI valve over-Fire in breaker 23-HOV-16.

Fire extinguished and breaker replaceJ.

760612 12 hrs, load or loss of control power.

Investigation revealed fire in BHCC 6.

76-25 760616 During ST-98, EDG full load test, it Problem in droop circuit.

Adjusted droop circuit and racked 760616 was noted that EDG's A, B, t. O were breaker in/out.

unstable in droop position from 09-8 4

panel.

76-37 760619 RCIC turbine would not trip due to Steam leak on 13-HOV-32.

Freed up and lubricated linkage. Reset 760619 rusty linkage.

and tripped linkage for approximatfly I week.

OR-76-90 760630 Failure of 02-RV-71E relief valve to unlinnwn.

Valve reset at approximately 500 psig 760630 i

reset following a turbine trip and coincidential with removal of fuses subsequent scram.

for 02-RV-71E solenoids.

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ll.K.3.ly REPORT Ott OufACES Of Et1ERGINCY LORE-COOLING SYSILMS LICtNSEE REPORT I.vi *s r DAll. lH:ltlkNils i f)llt he Ni llll/uRf I)Alli lilSCH I PI'lON CAtlSI.

tuimicg gyt:, Agg gggi m__._m_.

IU St.kvit:li lH)tvN IlHI -

76-65 761081 Fire in "A" Emergency Diesel Generator I.oss ol' Inlic ui t t o 'lurbucha rger.

Replaced turbo charger 761025 turbo-charger during ST-90.

76-67 761026 During ST-2H, valve 10-HOV-25A failed Notor burned up.

Manually opened valve 50 system would 761026 l

to open, funct ion LPCI avale - replaced motor.

?6-69 761808 While stroking 13-NOV-16, outboard Normal wear, anel torque suitch needed lhe inhoard valve was closed providing 16l101 RCIC isolation valve to verify limit adjustuwnt.

PCIS and 13-HOV-16 was manually opened switch operation, the valve bound up valve solor was replaced.

the autor operator.

k 16-72 761109 following scram with subsequent HPCI Apparent high i hne rate.

Started itPCI nunually and installed 161809 I I/2 his initiation,ItPCI started and then steam signal snubbing pins.

Isolated.

76-16 161115 During ISP-5-3 found pressure switch Ins t rimien t dr i f t.

Recalibrated immediately.

76til5 2-352C low out of Technical'$pecifica-tion requirements.

76-77 761117 During ST-98, "B" EDG failed to start.

Generator tach relays inoperable.

"A 6 C" EDG's sun with ECCS loads. The 76till 8 hrs, tach relay was replaced.

76-78 761187 During ST-98, fuel oil transfer pump Worn out.

Tested remaining pumps - OK and replace 1 761119 93 -P-C2 would not pump.

with spare pump.

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Enclasure 1 luger 20 I-II.K.3.17 Hl:ITIRT ON OlllM'.I.S OF IHi ktJ NCY Cultli-ftbl. LNG SYSTillS LICl.NSil: Rl:l'OHT f

I VI NI IIAIli Hl.1tlHHila I fpilth HI I I H/ Hit 8 18All!

DIld Hil* LION CalPAI'-

Cl3NIIICllVII ACIlOI4 10 St HVitT.

IN MdN 1158; n __

76-82 761122 Relief valve 02-RV-71E lif ted resulting Relief valve 02-RV-711: l i f t ed.

'the valve was rem >ved and replated 761123 in reactor scram from low pressure.

with a spare valve.

76-83 761122 Exceedeel maximum water volume in Relief valve 02-HV-7tE lif ted, l'uniped d<wn s'apression pool.

761122 4 lir.

supression chaniber after relief valve 02-RV-71E lif ted.

76-88 761201 loss of bellows monitoring on relief l'ound Isc groun t.

I.i f t e.1 ! cads unl m.ide 02-RV-71 A 761201 22 llrs.

valve 02-RV-71A.

inoperabig in the safety muJe fixed broken wire.

'6 93 7b1212 Hellows leakage alarm on relief valve Broken anephenol.

Iteplaced amphenol.

7bl232 02-RV-710.

76-9t 761215 During ST-98, "A" i:10 failed to start.

Rel i e f va l ve li f t ing on soa k bac k punip. "A" LIC was minually started and repeat 76l215 4

of ST-9B was successfully completed on next start (1ibe oli Iow pressure not in circuit for emergency start).

76-55 760821 During ST-24C, noticed st eam coming Rupture disc cracked.

Disc was replaced.

760821 15 Ifrs.

from leak-of f between RCIC s upture disc' 13-2-3, 76-63 761014 When shut t ing down "B" RilH inung, found Nut came of f disc.

clieck valve 42B leaking and unable to Replaced nut arul pegged.

761016 keep "B" EllR pump filled.

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,. Fo9e 21 ll.K.3.17 REPORT ON OUTAGI S OF IJil RGl.NCY Cole Ohll.ING SYSiltl$ l.lCl.NSI.E Ill:POR't IRNf IIAft! RIllllRNLD l~tplpill.tfr

. l.l R/OR8 DATE DESCHil"floN CAUSL CORRifflyn At' Tion TO St.RVICE InnrN IIHli

'77 06 770114 tharing routine surveillance, Ild "D" fl0 started on second start. No e<in i p-770114 I lir.

un low lobe oil pressure, tripped.

ment de fec t was found.

77-10 770224 turing routine surveillance,llPCI opes ator er ror.

Redundant switches were found operable.

77o3I1 level switches vt re found valved out l

of service.

4 77-17 770314 Iharing routine tests, RCIC turbine low oil level.

Auniliary oil sump was installcJ. Oil 770319 tripped while being started, was added tu turbine oil sump.

77-21 770120 foring routine surveillance test, Tachometer relay failed, lachometer relay was seplaced.

770425 FIC "A" failed to start.

77 28 770528 RCIC turbine damaged by overspeed.

Oil line failuie.

Site personnel lastalled a support bracke 770531 for oil line. IIPCI system was verified operable. RCIC turbine ove hauled.

77-29 770528 During normal operation, received Water seat in pressure sensing lines.

Lines were drained and Indicator checked 770528 2 lirs drywell hi h/ low pressure alarm.

with manometer. Installed drip legs on R

770621.

77-30 770528 During normal operat ion, a ground on DC ground.

Ground was repaired.

779916 02-RV-7tl. was discovered which During Wefuel prevented the bellows failure Outage indicators from working.

77-35 770611 lMring surveillance testing, 23-it>V-58 Unknown.

Valve was cycled and returned to service.

770hlt j

failed to open.

l 77-49 770928 lbring surveillance testing RCIC Improper sett ing of turnpic switth.

lor <tue switch was reset.

770930 steam supply valve 1341)V-131 would not open.

1 3

77 52 771003 During surveillance testing, ll!'Cl triuroper set t ing of tuivine swi t (li.

lor pne switch was reset.

771008 j

steam isulation would not fully close.

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_ __ur 77-59 771021 During ISP-474 a leak was discovered 1.ca k Pressure switch 02-PS-13tD was replace.l.

77ll17 in 02-PS-1340.

77-61 771116 Inst rument setpoint change for Operator error.

10-PS-120-A-Il were reset.

77ll17 10-PS.120A-Il was implemented prior t o NRC approval.

77 64-771123 hhile operating, received an alarm Ground on circuit. Ot er in pressure Cleared ground within a few days -

180225 indicating leaking bellows on 02-RV-71.1.

switch.

thought problem solved - in leh sleut domi unspected press switch and replaced.

77-67 77tl29 iMring ISP-9-1, inst rument channel Drift 773202 13-1)PIS-84 was found high out of 13-tipis-84 was calibrated.

ca1ibration.

77-71 771228 RCIC pump 13-P-1 w.as made inoperable Filter cloggeil.

ri t ter was cleaned.

771228 7 hrs.

for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to clean cuno filters.

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Fnclosure 1, age 24

'e

'll.K.3.17 REPORT ON OUTAGES OF EMLRCENCY CORE-COOL.ING SYSil.M'i LICLNSEE REPORT I

DAl'E RClijRNI.D rytJIPMLtf 1.LR/ORI DATE DESCRIPI'lON CAllSE CORRirflyE ACTION

'lO Sf.HVICE INNN Tith:

n i

78-70 780828 lauring normal operation, operator Pumn misalignment.

Aligned motor and pimp.

780828 5 t/2 hrs noted noisy operation of "A" fDG Iuhe oil soak liack pump causing "A" FDG to be plated out of service.

78-75 780902 During ST-3D, 14-FR)V-128 would not liigh differential pressure.

14 4k)V-128 was cycled manually, 780902 N/A open electrically, electrically and declared operable.

$ 78-81

'780915 During ST-211, circuit breaker for RilR Power fuse holder bent.

Straightened fuse holder clip.

790915 t/2 hr.

pimp 10-P-3C would not charge, rendering pump inoperable.

78-83 780923 1.ogic power for "B" LPCI subsystem Operator error.

Power restored.

780924 20 hrs de-eneigized for ADS maintenance.

78-97 781215 During ISP-10, eight temperature Operator error.

Temperature switches readjusted.

781215 2 hrs, switches used for llPCI line rupture were found improperly set.

78-98 781205 During start-up "A" EDG tripped while Misadjustment of engine governor.

Governor properly adjustest.

781205 3 hrs, being paralleled.78-109 781224 During start-up "A" RilR loop could not Discharge check valve failure.

"A" Rllit pumps were placed in operat ion 781225 38.8 hrs.

be maintained full of water.

in the Torus cooling mode to comply with Technic.at Specifications s

l a

.5

~..

~

.m 4

icnclosure 1, pagu 25 y

ll K.3.17 RiiPORT ON (MITAGES OF IJll:RGl_NCY COHE-C001.ING SYST1.tts l lCI NSI.l: Rt.POR 4

=

LVI:NT 933g; py;3 pap 3 g3 g.q,,g p,,, g7 LLit/OR8 DATIE DESCRIPilON cal 8Sli CORRICflVI! ACflON 10 SI:RVICE IMmN TlHI!

79-11 790234 During normal operat tun. Drywell Hlown l'use due to short in alve Iteplaced fuse and solenoid fuse blew 790234 2 hrs.

atmupheric sample isolation valve solenoid.

again on March 6,' 1970 and Septenher 790902 Valve cach time 27-SOV-123B failed.

2 1979.

Each time the other isolation was repaircJ valve in the line (27-SOV-12.5A) was satisfactorily place 1 in the closed position and sharing pipe verified shut daily.

stress outage.

!79-20' 790327 tharing ST-98, air compressor 93-AC-BI I)i r t in unloader.

Unioader was disassembled and cleaned.

790514 did not load.

Pise Stress l

Outage.

i

79-23 790ll!

During inspection of IIPCI turbine llPCI turlitne inspection.

None 790615 piping for fixed fire suppression.

plant in cold system, turbine must be disassembled shutdown for violating Technical Specifications.

pip-stress.

l l

79-24 790428 Emergency service water pump failed to Normal wear.

Overhaul and minor audification of 790M03 308 has.,

meet shutoff head requirements of impeller.

Plant in cold TS-4.!!.D.I.6.

shutdown for f

pipe stress.

i

79-35 790607 RCIC low pump suction pressme switch Out of calibration.

If,C personnel recalibratcJ pressure 790607 2 hrs.

setpoint too high (15.7" of mercury) switch.

Plant in vold Technical Specifications respaire shutdown for

<lsa pipe stress.

79-39 790627 thsring ST-35A, "B" loop itCl injection Failure of torque switch in motor Replaced torque switch assembly.

790702 valve failed to open (10-HOV-258).

operator.

pine stress outage.

s

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P

't I

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o A

I l

I:nclosure 1, gwge y, ll.K.3.17 REPORT ON (AITAGES OF I)!ERGENCY CORE-COOL.lik; SYSTIJIS 1lCl.r4SIE RI:IURI EVIRI DATli HElllitNI.H ftpfltt4 NT

?lIR/ORf DAIE Dl'SCRIPTION CAUSE CORRiff6VE ACT881N 10 StRVict

'00mN TIHl:

j79-40 790630 (Mring ST-2A reverse ritat ion of Discharge check valve not tui*y closeJ.

Valve disassenhied and repaired.

790630 i.

to-P-3A was noted.

79-42 790713 IMring ST-2R, RilR pumps could not meet Orifice in elinharge line.

Orifice in discharre line tempor.arily 790523 required Technical Specification t emoveil. afte rest ring hole.

Pipe stress flow rates.

outage.

79-46 790830 Deficiencies in the design ol' cable llPfl and N'S cabics in the same tray.

Rerouted flPCI and ADS cables with 800804 I week separation tetween ADS and alPCI terminat ion pos formed during refuel outage.

. 79-56 790903 IAsring plant heatup, the llPCI turhine Technical Specil scat ions require pirys Re.ictor pressure was maintained below 790903 N/A was intentionally uncoupled from the operable prior to st.artup - af ter 150 psig, f

pump (for overspeed test) violating commencement of startup, plant can do Technical Specification Appendix A, overspeed test.

3.5.C.I.a.

79-64 790906 During normal operation, while perform-Inadequat e comamicat ion.

2348)V-57 was placed in the open 790906 s hrs.

ing ST-4E the mitor on condensate position as soon as the improper storage tank suction valve failed.

condition was discovered.

The smitor from 23-ftW-57 was to be used to replace it.

Ilowever, when t he motor of 23 et)V-57 was removed the valve was not checked in the open position g

violating Technical Specifications.

.79-71 791001 During normal operation, emergency thalletin 79 02, u7, 14 ef fou t.

Hl5-124 was modified.

791007 service water pipe support 1115-124 was found inoperable.

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-.. - r --. _., ~,

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l e

4 Vuel oSut et k, 1*.ne 21 II.E.3.17 RI.PORl' ON eMI AGLS OF IH1 Rt.lhCY Colui-000 Lit 0 SYSil.tt.; 1.101 tGl.l. i<ll'unT I

IYIMI I)All kl.itildiH it IbilpHlWI l.[R/ORf DA'IE DESCRit' TIUM CAllSl!

CURRLCliVE ACTIOtl go si pygti; t.Mrt TlHa 79-73 790914 lauring ST-9H, l.LKi "C" t ripped on over-Inose connection on oversprest limit 1.ome connect ion was repa i red.

7909:4 speed during the engine start sequence, sw i tc h.

l79-81 791018 4 RilR pipe supports 1110-135, 1110-140,

!!ulletin 79-02, 07, f. Il effort -

Mala fic.a t ion to t he suppurt s was 791025 i

1810-168, and l'i:SK-72 7 found t o lie determination to repair rat her t han completed within t he 7 day t ime frame, r

inope rahic.

further analysis m.nle.

79-86 791025 6 pipe support s, Ik,6-62, lith I til6-2, Niificat ion au-l upgrade to t he 791101 ll4b-3, llW267A.1110-175, and 1110-177 supports was completcJ within the determined inoperable.

7 J1y time frame.

79-87 791029 3 RilR pipe support s determined lipgrading of the 3 support s was 79 1us I

inope rable.

completed within the 7 Jay time frame.

,79-89 791102 3 Suplmrt s in the Standby Liquid Contro Modifications and upgrading of the 791109 l>

system determined inoperahle.

till-8, suppor t s was rumpleted within the 7

]1 1166-33 and 1176-61.

Jay time frame.

I 79-91 791020 During ST-2R, Ritit pump "C" failed I.imit, switches for 10-l*N-l ic Li mi t switches for 10-MOV-13C were 791020 4 hrs.

to start properly, improperly adjusteil.

readjusted properly.

79-90

_91011 As a result of hanger mihlification Bullet in 79-02, n7 f. 14 ef fort.

Calite for 23-linV-17 was rerouted.

191014.

for pSA, the cable for 23-llOV-17 i

reiluired relocation disabling the I

valve. 1his caused the llPCI system to I,ccome a elegraded muite.

79-92 791013 During normal operation, an Loose lead en llPCI area temperature Connection was tightened.

791013 2 his.

intermittent isolation signal to switch.

23-HOV-15 was noted.

t l

.e t

..m t.nclosute 1, p.=9e 2d 1I.K.3.!7 RLPORT ON OlffAGES OF 111ERGl;NCY CORii-C001. Ital SYS'l17IS 1.ICl NSI.E Hi PORT l'VINF.

DATE RC1HENED IyllRNT -

I.LR/ORf DAll!

DESCRIPTION CAUSE CORRLCTl?E A(TION TO SI:RVICE IE'wN TIME I

1 79 97 791030 During normal oper tion while fligh resistance cont acts on relay Contacts on relay LSR-400 were cleaned.

791030 9 hrs.

performing ST-90,1.tx1 "C" would not ESR-400.

pa r t ie l wi t h 1:DG "A".

79-I00 791108 During normal operat ion, in order to I.eak in hyds.sulic cont rol valve sea t Gisket for hydrau'ie control valve 7911oa 4 hrs.

complete maintenance, the llPCI J ain g.isket.

seat drain was rept.sted.

3 system was intentionally made i

inope rabi c.

I 79-103 791205 RilR pipe hanger 1110-539 was determined Bulletin 79-02, 07, 14 etfort.

Modifhation to 1110-539 was completed.

791706 inoperable.

k 79-105 791114 During normal operation, LDG "C" was, had t> earing in motor.

"C" i.INI Hecirc lube oil pine motor was 791114 declared inoperable due to a nolsy replaced.

lube oil Recirc pump.

,75108 791206 During normal operation, the HCIC Could not I.ositively determine cause Welding in the area was finally determis ed 791208 48 hrs, system was declared inoperable to initially. Suspected excessive as the cause.

Inspect turbine bearings.due to bearing temperature. Decided to i

operator smelling smoke.

Inspect bc.aring tr he sure.79-110 791214 During normal operation, III'Cl steam 1.oose wire in a thermocouple switch.

The loose wire in the thermocouple was 791214 i

supply inboard isolation (23-PalV-15) retcrainated.

i closed as a result of a spurious isolation sigual.

. i 1

f

e

.,lege 29 11.E.3.17 RI: PORT ON OUTAGES OF iJtl RGLf4CY CORE-C001 Intl SYSTISG l.lCI:SI E 10'iOHT 1:VIRT DATE RCIOHNI D I t.Mll PHEN T l.LR/0R8 DALE Dl:SCRIl*fl0N

(:AUSE CURRfCTIVE ACI'lON jo 3ggyggg neh4 ygpg-

'80-27 800322 Ikaring normal operation, the HPCI Pipe hanger w.es repaired.

800322 s hrs.

i system was degraded to allow for pipe Bulletin 79 02, 14 ef fort.

hanger repair.

80-29 800325 During ISP-75, 23-LS-74A was found Drift.

Pemove, clean and calibrate 23 74 A.

800325 I hr.

set less conservative than allowed.

80-34 800410 tharing surveillance test, itPCI Cot ked brush holder.

Motor was replaced.

80.1410 8 hrs.\\

N minimum flow valve failed to close

, f80-35 800415 tharing normal operation, 10.p10v-67 Open torque bypass limit switch.

Aljusted timit n.hches.

Ed1415 8 hrs.

I failed to opeti.

t 80-38 800418 During normal operation, valve 10-HOV-Seat and disc wear.

Ove: hauled valve and refurbished sent 800610 4

l 57 was found incapable of fully and disc.

Re fue ling l

closing.

Outage.

80-40 '

800422

'During normal operation, the "A"

Cate timing card failures.

' Replaced rate timing card, filter 800123 5 hrs, f LPCI independent power supply capacitors and internal power supply, s

q inverter tripped.

g w,

j i.

l

' 80-47 800501 Iksring' normal operation, 13-TE-898 Broken wire.

4 Repaired broken evire.').

80050I to hrs.

f.miled to operate.

Mi

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,t N

i \\

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h.

C 80-06 8005]O..

.During normal shutdown maintenance;,

Dirty vultage regulator elroop switch.

Cleaned contacts un switch.

80t)610 EIC "C" tripped'on high circulatic.g Refuel Out ge

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c u rrc rit'.

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s

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i 80-62.

600717 1,uring normal shutdown, testing RilR Orifice in discharge, tine.

Orifice was removed.

800717 3 hrs.

x service water pumps A an! C l' low f,

was less than required by Technical i

e t

s Speci fic Ation 4.5. B.1.

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4 II.K.3.17 RiiPORT ON OlffAGliS OF l.MI.lu;11NCY CORii-COL 1.INti SYS11tlS l.ICl.NSEE REPOH f L. closure 1, pa.3e 30 t

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DATli RlilHRNiiD l-QUIPMENT LER/ORO DA1E DrSCRIPf!ON CAust cORRfCTIVli ACT!Ott TO SERVICE DowN TIME 80-04 800107 During surve'lliance testing, une of the Instrument Jritt.

Corrective action recalibration 800107 four core spray injection permissive immediately.

switches was found incorrectly set 02-3-PS-52A.

24 hrs.

80 06 800108 During a review of work requests, it Tnrque limit switches improperly Adjust torque and limit switches.

791221 was noted that core spray injection adjusted and en essive dif ferential valve "A" had not operated properly on pressure across valve.

December 21, 1979 (14-MOV-17A) 80-07 800109 During nosmal operation, "B" R11R loop 1110-28 had failed pull test required flanger 1110-28 was repaired.

800109 32 hrs.

was declared inoperable while hanger for Bulletin 79-02.

1110-28 was inoperahle.

80-08 800109 During ST-3D, "B" core spray loop Improper torque switch adjustment.

Readjust torque switch.

800109 2's hrs.

injection valve (14-MOV-12H) failed.

80-12 800111 During ST-98, EIC "A" tripped on low Partial failure of engine emers ten lleater was replaced.

80!!!3 48 hrs, tube oil pressure, heaier.

80-15

800123, During normal operation, 1.PCI Failure M filter cap citors.

Replaced two failed capacitors.

800124 9 hrs.

Independent power supply "A" inverter

tripped, j80-18_

800201 During smrmal operation, core spray Receipt of Technical Specification Recalibrnted detector to new Technical 800201 4 hrs, t

pipe break detector was outside thinage af ter rnluired impitmentation Specification valvo.

Technical Specification limit s in date.

table 3.2-2.

80-24 800213 liuring normal operation, llPCI steam Defective electrical connection.

Connector was reterminated (23 *III-92A).

800214 20 hrs.

l isolation closed as a result of a spurious signal (23.MOV-15).

4 1

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1. ' Enclosure 1, pa9e 31 y

li.K.3.17 REPORT ON 00fACliS OF IJt! RGENCY Ctmli-CODI.ltr. SYSil)l3 l.lClint:010. PORT

' ~

/

um DATE RElllRNED 1.Qt!!PHr.rtr -

{

LVINT lLER/ ORE DATE DESCRitrflON CADS!!

CORRI.Cf!VE ACTION TO SERVICE'

. DOWN TlHE a

1

e 4hr(/- '

80-66 800805' thiring ISP-8, core spray pipe break Drift.

Instiuments were calibrated and returne<

B00805 detection switches 14.DplS-43A anil to service.

d..

438 violated Technical Speci fication-table 3.2-2 limits.

l 80 67 800805 During ISP-12,llPCI and liCIC low Drift.

Instruments were calibrated.

800805 5 hr.

pressure witches 23-P3-68A and 13-.

5-878 were found set higher than that y

allowed by Technical Specification

~

Qt table 3.2-2.

80-69 800807 Dur ing ST-22A, - ADS t imer was set lower Drift.

Times was adjusted.

800308 3 hrs. P than t hat allowed by Technical

-]

7 Speelfication table 3.2-2.

80-71 800810 During plant startup, RCIC turbine railed amplifier Amplifier was replaced.

800810 2 hrs._

speed would not control in manual, 80-72 800812 During plant startup, "B" 1.PCI Electronic control failure due to one Causes of trips were repaircJ and the 8008I2 inverter tripped. Trips also tot 4 final control board.

central control board which created 4

place on 8/22/80, 9/5/80 and problem was finally identified and corrected.

9/6/80.

80

800913 During normal operat ion, "B" 1.PCI Failure of logic card A.

logic card was replaced.

800917 96 hrs.

2 invester tripped.

80-79 801008 During normal operation, RCIC system Normal wear.

Replaced valve body to inmnet gaskets.

801008 24 hrs, was made inoperable to allow repair of system drain vnive steam leaks.

80-80 801011 lharing normal operation, 10-MOV-57 thitor operator clutch was disengaged Removal of scaf folding in the area 801011 I hr.

would not close when required by due to scaffold installat ion.

. restored the valve to operable status.

Technical Specification 3.7.0.

T

i 9

11.K.3.21 RESTART OF CORE SPRAY AND LPCI SYSTEMS The Authority endorses the position taken by the GE/BWR Owners Group regarding this issue. This requirement has been reviewed by the GE/BWR Owners Group on a generic basis and it is concluded that the NRC suggestions are not required for plant safety considerations. This conclusion is based on the adequacy of the current ECCS logic design coupled with the potentially negative impact on overall safety of the proposed changes. These negative impacts include a significant escalation of control system complexity and restricted operator flexibility when dealing with anticipated events.

It is the Authority's understanding that the details of this analysis will be submitted through the GE/BWR Owners Group by January 1,1981.

The Authority does not plan to make any plant modifications regarding this issue., page 32

e ll.K.3.22 AUTOMATIC SWITCHOVER OF RCIC SYSTEM SUCTION The Authority has approved clear and cogent procedures at the James A. FitzPatrick Nuclear Power Plant for the manual switchover of the RCIC System suction from condensate storage tank to the suppression pool., page 33

r ll.K.3.27 PROVIDE COMMON REFERE!CE LEVEL FOR VESSEL LEVEL INSTRIJMENTATION The Authority, as part of the GE/BWR Owners Group, has reviewed the reactor water level instruments and has concluded that the current instru-mentation provides the FitzPatrick plant operators with reactor water level information that will permit them to make timely and correct decisions re-garding reactor water control requirements.

It is the Authority's under-standing that details of this review will be submitted through the Osiners Group by January 1, 1981.

It is the Authority's position that, on the basis of safety considerations, no modification of plant control room reactor water level instrumentation is required., page 34

t e

ll.K.3.44 ADEQUATE CORE COOLIDG FOR TRANSIENTS WIT 11 A SINGLE FAILURE An arialysis was performed through the GE/BWR Owners Crr.up to determine whether or not the reactor core remained covered during anticipated transients with a single failure. For a BWR/4 (J AFNFP), the worst case of transient with a single failure is a loss of feedwater with failure of the high pres-sure coolant injection (ifPCI) System.

It is concluded that the core remains covered during this worse case.

Addi tionally, it is concluded that for severely degraded transients beyond the design basis where it is assumed that an S/RV sticks open and an additional failure occurs, the core remains covered with proper operator action.

It is the Authority's understanding that further information regarding this issue will be provided by the GE/BWR Owners Group by January 1, 1981., page 35

ll.K.3.45 EVALUATION OF DEPRESSURIZATION WITH OTHER THAN AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)

Analyses of depressuri::ation rates other than full ADS were performed to determine the effect on reactor vessel integrity and core cooling capability.

It is shown that:

1.

Vessel integrity limits are not exceeded for full ADS blowdown.

2.

For slower depressurization rates, there is little impact on vessel fatigue usage relative to full ADS blowdown, and 3.

Slower depressurization rates have an adverse impact on core cooling capability.

Therefore, the Authority does not contemplate making any change in the ADS blowdown rate.

It is the Authority's understanding that further infor-mation regarding these analyses will be submitted to the NRC through the GE/BWR Owners Group by January 1, 1981.

J, page 36 a

0 lll.A.2.1 EMERGENCY PREPAREDNESS The Authority's upgraded Emergency Plan has been submitted under separate cover via the letter from J.

P.

Bayne (Power Authority) to D.

G. Eisenhut (NRC) dated Decerrber 30, 1980., page 37

_=. -

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.i i

j III.A.2.2.

METECR0 LOGICAL MEASURING SYSTEM Wind speed, wind direction and temperature sensors are installed on a suitable isolated tower at elevations of approximately 30 feet and 200 feet above plant grade. The data collected by these sensors are tele =etered to the JAFN?P control l

room'and are continuously recorded on strip charts.

This data will be available in the TSC and the near site EOF.

Monthly and semiannual joint frequency distributions of wind speed and direction 1

by atmospheric stability class are maintained to aid in the evaluation of potential l

radiation doses which may result from the release of radioactive material from the i

plant during an emergency.

The present system meets part of the new meteorological requirements specified in NUREG-0654, Rev. 1, Appendix 2 (Dated November 1980).

It ceets the requirements for l

a Class A :odel; that is, a model and calculational capability which can produce j

initial transport and diffusion estimates for the plumeEPZ within 15 minutes i

following the classification of an incident. A new system will be designed to r.eet the other requirements listed.

1 Catil the new system is installed backup meteorological data will be supplied by local weather stations. The national weather stations at Hancock Airport in Syracuse, N.Y. anc at the national weather station at Buffalo, N.Y. will provide access to their meteorological data, as required.

The new meteorological measuring system will meet the basic functions needed to comply with the radiological aspects of the 10CFR Part 50, 47 require =ents, namely:

I o

A capability for =aking meteorological =easurements.

A capability for =aking near real-time predictions of the atmospheric effluent o

transport and diffusion.

o A capability for remote interrogation of the atmospheric measurements and predictions by appropriate organizations.

it is expected the new system.will meet the requirements of NLT.EG-0654, Appendix 2 for technical requirements and implementation schedule.

Development of the new =eteorological measuring-system is being jointly undertaken by Niagara Mohawk Power Corporation and the New York State Power Authority.

1 i

-Enclosure 1, page-38

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O lll.D.3.3 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS The James A. FitzPatrick Nuclear Power Plant has the capability to provide iodine monitoring under accident conditions. Samples taken using sample media that will collect iodine selectively over xenon can be collected from accessible areas throughout the plant and counted for activity on a GeLi analyzer system in the plant laboratory. Should the regular counting facility be inoperable, due to high background activity, the samples can be analyzed at our on-site environmental laboratory, which is equipped with a filtered ventilation system to ensure low contamination levels., page 39

O III.D.3.4 CONTROL ROOM HABITABILITY The Authority is currently reviewing a draft evaluation on control room habitability.

It is expected that the required information will be submitted by January 30, 1981., page 40