ML18086A238
| ML18086A238 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/21/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18086A237 | List: |
| References | |
| NUDOCS 8104130172 | |
| Download: ML18086A238 (13) | |
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Enclosure l EVALUATION OF LICENSEE'S COMPLIANCE WITH CATEGORY 11A 11 ITEMS OF NRC RECOM!~ENDATIONS RESULTING FROM TMI~2 LESSONS LEARNED PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION UNIT NO. 1 Docket No. 50-272
1 -
Introduction Sy letters dated October 12 (Reference 1), November 23 (Reference 2),
December 14 (Reference 3), December 14, 1979 (Reference 4), January 2 (Refer~nce 5), April 11 (Reference 6), April 14 (Reference 7), and July 8, 1980 (Reference 8) and March 2, 1981 (Reference 9), Public Service Electric and Gas Company (licensee) submitted documentation of actfons CO;tii)leted and commitments of actions to be taken at Salem Nuclear Generating Station, Unit No. 1, to implement staff requirements (References 9 and 10) resulting from TMI-2 Lessons Learned.
During this time frame the licensee was providing identical or similar information related to Salem, Unit No.
- 2.
At the conclusion of our review of the licensee's input through January 2, 1980, the staff met with representatives of the licensee at the Salem site on April 8-9, 1980, to discuss action items that had not been fully approved.
Subsequent to this visit, the licensee provided the additional information and justification that had been requested.
The follm*ling is our evaluation of the actions taken by the licensee to ir::p1 ement each Category "A" item by the scheduled deadline of January 1, 1980.
Ev al ua tion Each of the Category 11A" requirements applicable to PWRs is identified below.
The staff's requirements are set forth in Reference 10; the acceptance criteria is documented in Reference 11.
The numbered designation of each item is consistent with the identifications used in NUREG-0578.
Lessons Learned items 2.l.7a and 2.1.9 ~re being reviewed separately and are not discussed in this report.
- 2. 1. 1 EMERGENCY POWER SUPPLY Pressurizer Heater The Westinghouse 0.'lner's Group analysis has determined that the minimum requirements to maintain subcooling, in a four loop plant with a pressurizer volume of 1800 cubic. fee~ is 150kw of heater capacity.
Two backup heater groups, each rated at 400 kw, have the capability of being manually connected to redundant diesel generators upon loss of offsite power.
Tne* diesel generators are capable of supplyjng the 150 kw of pressurizer heaters concurrent with t.he equipment loads required for a LOCA.
The onsite emergency*powe1 system for unit one includes three 2600 kw.
diesel generate~ and three 125 volt batterie$.
Emergency Procedure EI 4. 9, 1181 ackout" address es the trans fer of the heaters to the vital buses.
Under blackout conditions, the diesel generators have sufficient capacity to supply the required equipment loads including the pressurizers heaters and meet.the continuous diesel rating.
The connection of the pressurizer heaters to a vital bus is through a normally open Class IE circuit breaker.
The setup of the vital bus feed to the pressurizer heaters can be completed in a time frame consistent with maintenance of natural circulation.
- 2. 1. 2 2 -
The licensee has satisfied the short term lessons learned requirements for pressurizer heaters.
-Pressurizer Relief and Block Valves and Pressurizer Level Indicators T1*10 power-operated relief valves (PORV's) for the pressurizer are pneumatically operated* from the instrument air system upon actuation of solenoid control valves.
The instrument air system is supplied from three station air compressors.
An instrument air accumulato'r for each PORV located in the containment provides approximately '100 stored valve operations for each valve.
Two instrument air compressors energized from redundant emergency buses supply th~ instrument air system upon loss of offsite power.
The solenoid valves for the two PORVs are energized from separate 125 volt plant batteries. The block valves (motor operated) for the PORVs are energized from redundant emergency 230V AC buses which are fed from diesel generators A and B automatically upon 1 oss of offsi te pov1er.
The PORVs and their associated block valves are connected to the emergency sources of power through safety grade circuit breakers.
The design of the PORVs and block valves are such that they can be opened in addition to being closed in the event of 1 ass of offs ite power.
Three pressurizer level transmitter instrument channels indicate level in the control room on three sector gages.
Each of these level instrument channels is independently powered from a different vital instrument buses.
Each of these buses is inverter fed from a separate 125-V plant battery.
Tile licensee has satisfied the short term lessons learned require-ments of emergency power supplies for the pressurizer power-operated relief valves/block valves and pressurizer level indicators.
PERFORMANCE TESTING FOR PWR RELIEF AND SAFETY VALVES All PWR licensees are required to functionally test reactor coolant system relief and safety va.lves to demonstrate operability under expected operating conditi'ons.
The Category 11A 11 requirement is for the licensee to cpmmit to perform an appropriate test program.'
The licensee has referenced the Electric Power Research Institute's (EPRI) "Program Plan for the Performance Verification of PWR Safety/
Relief Valves and Systems," as the program description and schedule to meet this requirement. This action is acceptable.
- 2. l. 3. a 2.1.3.b*
2.1.4 DIRECT INDICATION OF POHER-OPER.~TED RELIEF VALVES AND SAFETY Vf>.LVES FOR PWRs NUREG-0578 requires PWR licensees to provide positive position indication for reactor coolant system relief and safety valves.
The licensee has installed seismic and environmental qualified limit switches, powered from vital buses, on both the power-operated relief valves and the safety valves. These limit switches indicate position and alarm on open positions and meet our requirements for this action item.
INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING The licensee is using the plant computer, which is powered by a vital bus, to display continuously the margin of subcooling on a control board meter.
The saturation temperature calculated using the average of tl*m vJide range, safety-grade, pressure sensors is compared with the highest of 65 thermocouple readings in the core to obtain the margin of subcooling.
The licensee is in compliance with this action item.
OIE will verify that the procedures to manually determine subcooling conditions in the core are adequate.
CONTAINMENT ISOLATION Three problems ass*ociated with containment isolation 1*1ere high-lighted by the TMI-2 accident:
(a) the need for diverse parameters for initiation of containment isolation; (b) consistent definition of essential and non-essential systems; and (c) the possibility that valves may reset to an incorrect position when an isolation signal is reset.
To counteract these problems, the staff has developed four positions to assure containment isola~ion.
- 1. All containment isolation ~yste~ designs shall comply with the recomoendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the i ni ti ation of containment isolation.
By letter of ~anuary 2', 1980, the licensee identified diverse signals for initiation of Containment Isolation Phase A (5 parameters), Containment Isolation Phase B (2 parameters),
Containment Ventilation Isolation (8 parameters), Main Steam Isolation (3 parameters), and Feedwater Isolation (8 parameters).
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- 2.
All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identi,fy each system to be non-essential, shall describe tr.e basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report to the results of the reevaluation to the NRC.
In its letter of July 8, 1980, the licensee sta.tes that Sa1em 1 conforms with the essential/non-essential categories and recommended isolation provisions of a report prepared by the Westinghouse Owners Group.
The essential systems consist of Residual Heat Removal, Safety Injection, Containment Fan Coolers-Service Water; Steam Supply to AFW Purr.p Turbine, Main Steam Atmosphere Relief, Auxiliary Feedi*Jater, and Cha.rging.
The licensee states that, as the result of its review of the systems listed in Table 5.4-1 and Figures 5.4-1 through 5.4-27 of the Salem Unit No. 1 FSAR no further changes in design are needed.
- 3.
All non-essential systems shall be automatically; isolated by the containment isolation signal.
All non-essential systems are either isolated up on containment A or B isolation signals, or have a non-return check valve or are closed and under administrative control during operation of the plant.
- 4.
The design of control systems for automatic containment isolation valves shall be such that resetting the isofaticrn signal 1*1il1 not result in the automatic reopening of contairn~ent isolation valves.
Reopening of containment isolation valves shall require deliberate operator action.
The licensee has revie1*1ed the isolation valve control systems to verify that the valves remain closed upon resetting the isolation signal until they are*deliberately reset by the operator.
Two systems were identified that required further design change to prevent inadvertent transfer of radioactive material from containment.
These 1*/ere the Reactor Cool ant Ora in Tank pump discharge line and the Press*urizer Relief Tank gas analyzer 1 i ne*.
- The llcensee committed to comp1 ete these changes on the Categofy 11A 11 implementation schedule.
Salem Unit No. 1 does not have any valve control switches that control the reopening of more than one valve; i.e., all valves are individually operated.
2.1.5.a
- 2. 1. 5. c 2.1.6.a
- On the basis of the licensee's review and responses we find his implementation of this actio~ item to be acceptable.
DEDICATED HYDROGEN CONTROL PENETRATION This requirement does not apply to Salem Unit No. 1 because there are redundant hydrogen recombi ners 1 oca ted v1ithi n containment.
RECOMBINER PROCEDURES The initial combiners are operated from the control room, consequently shielding of personnel and exposure 1 imi ta ti ans a re not of concern.
The 1 i cens ee has reviewed the procedures for using the recombiner in accordance v1ith the guidelines set forth
)n IE Bulletin 79-06A.
The hydrogen control system thereby meets the requirements for this action item.
INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT The licensee described, in its letter dated January l, 1980, a comprehensive leak reduction program for systems outside containment.
The details of this program were reviewed during the staff's visit in April 1980 and \\~ere partially revised in the licensee's letter of July 8, 1980.
The program includes all systems needed for operation of safety systems or needed to cool the core (e.g., SI, CS, RHR, WGS, liquid rad\\vaste, sampling, and CVCS).
The licensee also identified systems that were excluded from this-program and provided the justification for this action.
Leak measurements that could not be made while the plant is operating were completed during the refueling outage in October-November 1980 and verified by OI&E.
The results \\vere reported to the NRC by letter dated March 2, 1981, and found to be acceptable.
A preventive maintenance program has been established and includes regular surveillance for leaks, leak tests each refueling cycle and repair df detected leaks as soon as practical. Specific leakage criteria have not been developed, however, the licensee is committed to maintaining leakage as low as practical. Helium leak testing has not been planned;. therefore, special training or equipment is not require.
- 2. l. 6. b
- 2. 1. 7. b
- Design Review of Plant Shielding and Environmental Qualification A design review was conducted by the PSE&G engineering staff.
The NRC-specified source terms (with acceptable minor modifications)
\\*tere used.
Systems which are designed to function after an accident were considered as radiation sources (i.e., SI, RHR, Charging, RC &
Seal water filters).
Vital areas were identified and their post-accident accessability 1*1as evaluated using the dose limits of General Design Criteria 19.
AdditionEl shielding was found necessary for the charging pump valve compartment, the primary sample laboratory, the gaseous v1aste control center and the liquid waste control center.
The control room and the technical support center were found sui tab 1 e for continuous occupancy.*
The eva 1 uati on of radiological environmental qualification of safety equipment is continuing; the dose calculations have been completed but environmental qualification information is proving difficult to obtain.
Salem committed to completion of all modifications needed to ensure environmental qualification of safety equipment by January 1, 1981.
Sal em is in compliance 1*dth the intent of these Lessons Learned requirements.
/1.uxiliary Feedwater Flow Indication to Steam Generators The auxi 1 i a ry feedwater fl ow sys terns, one for each of the four steam generators, indicate flow in the control room.
Tv:o of these instruments channel~ are powered from vital instrument bus A and the other two instrument channels are pov1ered from instrument bus B.
Three narrow range steam generator \\*tater level i-nstruments (energized from separate vital instrument buses) for each steam generator readout in the control room to provide diversity and satisfy the single failure criterion.
The design of the auxiliary feedv1ater fl0\\*1 indication channels is such that on-line testing is a feature.
The feedwater flow indications to each steam generator are tested during startup and shutdown, and calibrated every three years.
Each auxiliary feedl':ater flOlv channel provides an indication of feed flow \\*1ith an accuracy of_:::. 2.7%.
The licensee has sa.tis'fied' the short term Lessons Learned requirements for auxiliary feedwater flow indication to steam generators for PWRs.
- 2. 1. 8. b
- Improved Post-Accident Sampling Capability A* design and operational review of containment atmosphere and reactor coolant sampling systems has been conducted.
Improved post-accident sa:npling methods will be developed.
The licensee's commi.tment to implement improved methods by May 1, 1980, \\'taS fulfilled.
TI1ese interim modifications consist of extending sample lines to new sa~pling locations, providing added shielding and taking sr.1al1 (microliter) samples.
S:;.mples ca;1 be to.ken frmii both the reactor coolant and the containment atm::1sphere.
Samp1e collection activities wi l1 not be 1 imi ted by radiation exposure so the s amp1 es can be obtained 1*1ithin one hour.
The containment atmosphere sample can be analyzed*f9r radioisotopic composition and for hydrogen content.
The reactor coolant sample can be analyzed for radio-isotopic composition, for boron concentration and for pH.
The Salem counting facilities are expected to be functional after an accident but a 101*1-background backup facility also is being provided.
Salem will be in compliance with these Lessons Learned requ i rementS.
OI&E will review the sampling and sample analysis procedures and vJill verify that the modifications of the sampling system are completed as scheduled.
Increased Range of Radiation Monitors Interim methods for monitoring high-level releases have been developed and are being implemented.
At Salem, most rele~ses would be through the plant vent stack so all potential releases are covered by monitoring the pl ant vent stack, the auxiliary feed1-Jater vent and the main steam line discharges.
The plant vent is monitored for noble gases \\-Jith a fixed area radiation monitor (a GM-type, Tracer lab TA-62) with a readout in the control room.
Procedures are in effect for converting the dose rate readings to release rates.
The plant vent is monitored for iodine and particulate releases with a sampling system.
Tneeffluent sample is drai*m through a filter paper and a silver zeolite cartridge~ The filter media are removed and analyzed in a low-background area using a multichannel analyzer or a SAM-II unit.
Procedures are in effect for handling and analyzing high level effluents.
The main steam* line discharges and the auxiliary feedwater vent discharges a*re monitored \\-Jith portable gam11a radiation mqnitoring instruments.
The instrument is used to determine the dose rate at a predetermined location.
Procedures are in effect for. converting the dose rate reading to release rate.
Tnese systems meet the power supply, range and reading frequency requirements.
Salem is in compliance with these Lessons. Learned requirements.
2.1.8,c
- 2. l. 9
- 2. 2.1. a
- OI&E will review the interim effluent monitoring procedures for adequacy, vdll verify that personnel have been appropriately trained in the use of these procedures, and 1<1ill verify that the effluent monitoring equipment is installed.
~roved In-Plant Iodine Monitoring Air monitoring at Salem is performed v1ith portable air sar:1plers.
The filt::!r media are removed and counted ~n c.n apprc;:;riate lO\\'I background area with SAM-II Units; 5 such units are available, thus qui ck results and a 1 ow background area are assured.
To pro vi de added capability for discrimination against noble gases, silver zeolite cartridges are available.
Personnel have. been appropriately trained.
Thus the capability exists for c.ccurately monitoring iodine in the presence of noble gases.
Salem is in compliance with these Lessons Learned requirements.
OI&E \\'Jill review the air monitoring procedures and verify that silver zeolite cartridges are SAM-II units are available.
Analysis of Design and Off-Normal Transients and Accidents -
Reactor Coolant System Vents The licensee has provided the design for the reactor coolant system vent and has addressed all of the clarification items in the October 30, 1979 letter.
This response has been reviewed and the design found to be acceptable.
Shift Supervisor's Responsibilities In response to concerns that the Shift Supervisor might be distracted from his primary management duties the licensee has taken the follovling actions to es tab 1 is h the authority of the Shi ft Supervisor.
A l'ffitten directive signed by the General Manager Electric Production, has been issued to describe and emphasize the responsibilities and command duties of the Shift Supervisors.
Administrative Procedure No. 5, "Operating Procedure, 11 has been revised to further.clarify these responsibilities-and.to delineate th*e command decision authority of the Se:nior Sli'ift Supervisor. Shift Supervisor's administrative fuhctions that detract from or are subordinate to the management res pons ib i 1 ity for assuring safe operations of the pl ant are delegated to other operations personnel not on duty in the Control Room or to other station personnel.
Training programs that emphasize the Shift Supervisor's management functions have been established.
The license has responded to this action itlm in an acceptable manner.
2.2.l.b 2.2.l.c 2.2.2.b
The 1 i censee has established the posit ion of the STA \\*1ho has the dual responsibilities of an evaluation engineer during normal plant operation and an advisor to the Senior Shift Supervisor during nuclear plant transients.
The STA reports to the station Reactor Engineer; however, under normal operation conditions he is under the functional supervision of the Shift Supervisor.
STAs have a bachelors degree or equivalent.
During 1980,interim STAs served on shift after ha*1ing received special training.
Fully trained STAs, 1*1ho had undergone 35 weeks of specialized training, were assigned as of January l, 1981.
The licensee is thereby in compliance with this action item.
Shift and Relief Turnover Procedures The licensee has implemented this action item as follows:
Check 1 is ts of critical plant parameters are provided in the operating log~ for offgoing and oncoming shifts.
The Ad~inistrative Procedures and Operating Department Manual have been revised tJ provide more formalized shift turnover procedures and to ensure.hat an adequate review of the shift logs are made.
The licensee o has a monthly management inspection of shift operations to ve**
operator under-standing of equipment status and plant alarms.
ddition, a tagging request verification provides an independent ch:.:.
.n system lineup and documentation of lineup.
The licensee is in compliance with this item.
Onsite Technical Support Center The interim onsite Technical Support center (TSC) has been established in the clean facility \\'Jithin the plant security boundary, located adjacent to Unit :*io.l. This area is readily available to the document center (one floor down) where pertinent plant records and plant dra1\\fings are available.
The emergency procedures provide plans for activating engineering/mariagement support and staffing of the TSC.
Dedicated corrmunicatidn bet~*1een the TSC,. the Control Room and the NRC Operation Cent~r in Bethesda, has been installed and is operational.
Radiation equipment for monitoring airborne contamination and direct radiation is located in the TSC.
i*!hile the Center is activated, it will be continuously monitored for airborne contamination and direct radiation.
2.2.2.c
- A slave CRT to the operators CRT in the Control Room is used to display plant operating parameters in the TSC.
Jl.pproxirnately two dozen plant parameters are available on this CRT.
A computer terminal also available in the TSC can call out of unit one computer any of approximately 500 pl ant opera ting pararr.eters.
Procedures have been implemented to provide plans for performing accident assessment functions in the TSC and* the Control Room or nearby Senior Shift Supervisor 1s office should the interim TSC become uninhabitable.
The licensee has satisfied the short terms Lessons Learned require-ments for ans ite Technical Support Center.
Onsite Operational Support Center The licensee has designated the enclosed area between Unit No.
and Unit No. 2 Control Rooms and the cafeteria as the on-site Opera ti ona 1 Support center.
Communi cations 1 equipment bet1*Jeen these areas and the Control Room presently exist. A procedure which describes the activation, manning, and use of the operational Support Center has been implemented.
The licensee has satisfied the short terms Lessons Learned require-ments for the onsite Operational support Center.
REFERENCES
- 1. Letter, PSE&G (Librizzi) to NRC (Eisenhut), October 12, 1979
- 2. Letter, PSE&G (Librizzi) to NRC (Eisenhut), November 23, 1979
- 3. Letter, PSE&G (Librizzi) to NRC (Denton), December 14, 1979
- 4. Letter, PSE&G (Librizzi) to NRC (Schwencer), December 14, 1979
- 5. Letter, PSE&G (Librizzi) to NRC (Eisenhut), January 2, 1980, 19
- 6. Letter, PSE&G (Librizzi) to NRC (Sch\\'1encer), April 11, 1980
- 7. Letter, PSE&G (Librizzi) to NRC (Schwencer), April 14, 1980
- 8. Letter, PSE&G (Librizzi) to NRC (Denton, July 8, 1980
- 9. Letter, PSE&G (Mittl) to NRC (Varga) March 2, 1981
- 10. Letter, NRC (Eisenhut) to PSE&G (Librizzi), September 13, 1979
- 11. Letter, NRC (Denton) to PSE&G (Librizzi), October 30, 1979
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