ML19210D301
| ML19210D301 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/21/1979 |
| From: | Kelly R GEORGIA POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7911260277 | |
| Download: ML19210D301 (7) | |
Text
.T Gecrga Power Corrpany 230 Peachtree S:reet Post Cifice Box 4545 At:arta. Georpa 30302 Te!ecnone 404 522 6060 R. J. Kelly Gyg( y PO\\\\'Cf Vice Pres: cert and Gereral Manager November 21, 1979 Dower Generation 3 % wm, g,rc me Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDWlN I. HATCH NUCLEAR PLANT UNITS 1, 2 AMPLIFICATION OF COR!ITMENTS REGARDING THE SHORT-TERM RECOMMENDATIONS OF TMI-2 LESSONS LEARNED TASK FORCE Gentlemen:
After receipt of your letter dated October 30, 1979, Georgia Power Company's submittal was reviewed with regard to its complete agreement with the staff's requirements as previously documented in NUREG-0578 and the September 13, 1979, letter.
In response to your letter of October 30, 1979, Georgia Power Company submits Short-Term Lessons Learned Revised Commitments Enclosure (1).
As additional information is supplied by the Division of Operating Reactors, or as further study by Georgia Power Company requires, the conclitments contained in the enclosure may be amended.
If problems develop such as delays in the procurement of equipment which would delay installation, a revised schedule for installation will be submitted.
The equipment qualifications will meet the requirements of the currently approved revision of Regulatory Guide 1.97 available at the time the equipment is ordered.
Very truly yours, R. J. Kelly WEB /YB/mb Enclosure xc:
Mr. Ruble A. Thomas George F. Trowbridge, Esquire 1300 156
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TB GO N P
6
7 2.1.2 Relief and Safety Valve Test Georgia Power Company is investigating: actions required to upgrade RFPT, HPCI, and RCIC trip on High RPV Level, to redundant / safety grade; requirements to analyze SRV's and tailpipes for two-phase flow, and evaluate impact on pressure switches, Mark I containment and piping / supports.
Procedures for preventing overfilling, stuck-open relief valves are being reviewed.
No testing is planned at this time.
2.1.3.A Direct Position Indication of S/RV Plant Hatch presently utilizes pre.ssure switches.
The existing Safety / Relief Valve Position Indication System consists of a pressure activated switch located on the tailpipe of each Safety / Relief Valve. The pressure switch energizes two relays. The first relay signals the plant computer each time the contact closes or opens, which in turn records the valve number, valve position and time of each event. The second relay seals in, annunciates in the main control room and lights a light in the main control room above the individual valve control switch.
The seal in feature of the second relay can only be reset by a kay operated switch. Power for the relays and alarms are provided from the station battery system for reliability of operation.
This system meets all the requirements of the BWR Owner's Group position, with the exception of documentation of qualification for the pressure switches. Vendor catalog information indicates the existing switches are qualified for temperatures up to 1800F and have operated satisfactorily in a SRV blowdown environment. However, if qualification documentation is not available for the existing switches, they will be replaced with suitable hardware during the first maintenance / refueling outage after they become available.
2.1.3.8 Instrumentation for Inadequate Core Cooling Because the BWR operates in all power generation modas with both liquid and steam in the reactor pressure vessel, saturation conditions are always maintained irrespective of system pressure. Thus there is no need for a subcooling meter in the BWR.
GE analysis on Phase II B & 0 question will be submitted to the NRC by 11/30/79 as scheduled.
Instrumentation will be reviewed per GE analysis results.
Procedures are being reviewed for use with currently existing instrumen-ta tion. A description of this instrumentation will be provided by 1/1/80.
2.1.4 Containment Isolation Essential and non-essential systems have been identified-definition and justification are provided.
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2.1.4 Containment Isolation (Cont'd)
Automatic valve reopening will be disabled by engineering design; plant modifications and operator porcedures will be implemented as temporary measure by 1/1/80. Pennanent logic modifications will be implemented when hardware is available.
Procurement of hardware is underway but not expected to be complete by 1/1/80.
2.1.5.A Dedicated Hp Control Penetrations To assure that containment integrity is not potentially impaired due to buildup of combustible gases following a LOCA, Edwin I. Hatch Nuclear Plant, Unit-1 is equipped with the containment atmosphere dilution (CAD) system and a containment purging system.
As is shown on the attached HNP 1 H2 CONTROL pef!ETRATION SCHEMATIC, (sce Figure 1), although eacn penetration has multiple uses they are each arranged in a single-failure-proof configuration. This configuration satisfies the concern addressed by NUREG-0578.
Edwin I. Hatch Nuclear Plant, Unit-2 is equipped with redundant and independent combustible gas control systems (CGCS) consisting of two 100%
capacity thermal hydrogen recombiners (see FSAR section 6.2.5).
The two thermal recombiners are located in the reactor building as shown in FSAR Figure 3.8.29.
Two separate penetrations (X-61 A for system "A" and X-15 for system "B") are provided for gas flow fron he drywell to the recombiner system, and two separate penetrations () alA for system "A" and S-2228 for system "B") are provided for gas ar.c water return from the recombiner system to the torus.
Each combustible gas control system (CGCS) line is provided with two normally closed valves in series for isolation purposes.
This configuration satisfies the concern addressed by NUREG-0578.
2.1.5.C Recombiners Not applicable to Unit 1 as it does not utilize a hydrogen recombiner.
Shielding review for Unit 2 is being conducted per 2.1.6.B requirement.
Procedures are being reviewed and will be revised as necessary.
2.1.6.A Integrity of Systems Outside Containment Likely to Contain Radioactive Material System identification and test procedures have been developed.
Testing is underway, but due to extensive nature, not expected to be completed by 1/1/80.
A report on results of completed testing, and completion schedule will be submitted by 1/1/80.
Maintenance program will be developed by olant and installed by 1/1/80.
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2.1.6.B Design Review of Plant Shielding Identification of systems and locations have been made and scope of study has been determined.
Equipment dose calculations will be reviewed to assess whether safety equip-ment will be unduly degraded by radiation fields during post-accident operation of these systems.
Description of interim options and permanent modifications proposed will be completed and sumbitted by 1/1/80.
2.1.8.A Post Accident Sampling Capability Interim modificationsfor temporary shielding are expected.
Procedures will be revised, as necessary, and implemented by 1/1/80.
Proposed modifications will be submitted for NRC review in 1980, for approval and implementation by 1/1/81.
2.1.E.8,C Interim Procedures for Qualifying High Level Accidental Radioactivity Releases (This response applies to items 2.1.8.B B1, B2, B3, and c)
Procedures for interim use of portable instrumentation are being developed.
Permanent hardware will be procured and installed when avail able.
Proposals and schedules will be submitted to NRC in 1980 for implementation by 1/1/81.
Procedures will be modified, as necessary, to meet the needs of the new hardware.
2.1.9 Transient and Accident Analysis GE will submit guidelines and analysis to NRC by 1/15/80 as scheduled.
Georgia Power Company will implement the guidelines as appropriate when available.
ACRSl Containment Pressure Indication Pressure transmitters (-5 to +225 psig) for primary containment pressures will be added.
Proposed plant modifications will be submitted to NRC in 1980; for imple-mentaion by 1/1/81.
ACRS2 Containment Water Level Indication A review of water level indication system shows that the system meets the intent of the requirements, although not in complete agreement with the staff's position as stated.
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ACRS2 Containment Water Level Indication (Cont'd)
Edwin I. Hatch Units are equipped with a torus water level monitoring system capable indicating level in the control room from 21h" above the torus bottom to 11' - 7h" above normal water level.
Since normal torus water level is 148" above the torus bottom and a water level dropping below 51" from the torus bottom begins to uncover the ECCS suction lines, level indication below 21h" is no longer relevant.
This system meets the intent of Regulatory Guide 1.97 because it is redundent and safety grade.
ACRS3 Containment Hydrogen Indication A review of the hydrogen indication system shows that the intent of the requirement for this item is met.
Unit 1 is equipped with dual range hydrogen monitors, 0 - 10% and 0 -
100% concentration and is continuously recorded in the control room.
This system meets the intent of Regulatory Guide 1.97 because it is redundant and safety grade.
Unit 2 is equipped with dual range hydrogen monitors, 0 - 5% and 0 -
50". concentration and is continuously recordei in the control room. This system meets the intent of Regulatory Guide 1.37 because it is redundant and safety grade.
ACRS4 Reactor Coolant System Venting Georgia Power Company concurs with the Owner's Group position, as follows:
(1) The Owner's Group believes that adeqate reactor coolant system venting is provided by the existing plant design.
(2) Plant procedures will be provided to govern the operator's use of the relief valves for venting the reactor pressure vessel.
(3) No new 10CFR 50.46 conformance calculations or containment combustible gas concentration calculations are required, since systems in the plant's original design and covered by the original design bases are used.
2.2.1.A Shift Supervisor Responsibilities Georgia Power Company concurs with the BWR Owner's Group implementation criteria and will resolve this requrest accordingly.
2.2.1.B Shift Technical Advisor SR0's, who will perfonn normal operational duties under routine con-ditions, but will be removed from the chain of command in accident con-ditions, will be given limited engineering training and will be installed in this position by 1/1/80 to perform interim STA functions. Operational assessment functions will be performed by assigned technical staff engineers by 1/1/80.
In 1980, engineers will be given operator's training to take permanent STA positions and operational assessment functions by 1/1/81.
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2.2.1.C Shift and Relief Turnover Procedures Following the accident at TMI-2, a review was conducted of the shift and relief turnover procedures at Plant Hatch.
It is also Georgia Power Company's intent to implement the BWR Owner's Group Criteria.
2.2.2.A Control Room Access Georgia Power Company concurs with the BWR Owner's Group implementa-tion criteria and will resolve this request accordingly.
2.2.2.B Onsite Technical Support Center Georgia Power Company concurs with the BWR Owner's Group implementa-tion criteria and will resolve this request accordingly.
2.2.2.C Onsite Operational Support Center Georgia Power Company concurs with the BWR Owner's Group implementa-tion criteria and will resolve this request accordingly.
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