ML20214X350
| ML20214X350 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/01/1986 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-82-33, OL-1-I-SAPL-001, OL-1-I-SAPL-1, SBN-499, NUDOCS 8612110077 | |
| Download: ML20214X350 (13) | |
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' Q C{ N { D'.F. 37.1.2 V2.2.1 United States Nuclear Regu*SotMYCor3miPsIb Washington, D. C. 20555
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Attention:
Mr. D. G.PMe,,nhut, Dire ~c't 6/ '.
Division of Licensing Office of Nuclear Reactor Regulation Referenc es:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 gn, (b) USNRC Letter, dated December 17, 192, " Supplement 1 to
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NUREG-0737 - Requirements for Emergency Response l
Capability (Generic Letter No. 82-33)," D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits
Subject:
Response to Generic Letter 82-33; Supplement 1 to NUREG-0737
Dear Sir:
In response to the referenced letter in which you requested information pursuant to 10CFR50.54(f), we have enclosed a discussion of the status and have included commitments for implementation and integration of each of the following NUREG-0737, Supplement 1 items:
Emergency Response Facilities (includes meteorological monitoring)
Accident Monitoring Instrumentation (Regulatory Guide 1.97, Rev. 2)
Emergency Operating Procedures Detailed Control Room Design Review
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Safety Parameter Display System O
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Please contact me should you require additional information.
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Very truly yours, g[
8612110077 861001 g.
PDR ADOCK 05000443 YANKEE ATOMIC ELECTRIC COMPANY f
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.h.DeVincentis Project Manager ALL/fsf Enc losure cc:
Mr. George W. Knighton, Chief Mr. Louis Wheeler, Project Manager Licensing Branch 3 Licensing Branch 3 Division of Licensing Division of Licensing Atomic Jafety and Licensing Board Service List
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Following the declaration of an emergency, the activities of the emergency organization will te coordinated in a number of deditated Emergency Response the shift personrel Facilities.
During the initial stages of an emergency, initial emergency classify and declare the emergency condition and directThese actions include no
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of on-site and of f-site emergency response personnel, corrective actions, dose response actions.
As additional support assessment and protective action recommendations.
personnel arrive, certain emergency response functions are transferred to Center, Operational Support Center, personnel located in the Technical SupportFor information and details about and the Emergency Operations Facility.
meteorological monitoring and the use of meteorological parameters for see the PSNH letter to atmospheric dispersion and dose projection purposes,18, 1983 ("Open Item Respon the NRC, dated January G. W. Knighton).
Technical Support Center On-site Technical Support Centers (TSC), located in each units' control Building, have been established where assigned personnel diagnose accident
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conditions and assist the Control Room staff in plant stabilization (see The location and communications equipment of the TSCs facilitate voice and data interaction and coordination with the Control Room and th Figure 1).
Each TSC is included in the station la Emergency Operations Facility (EOF). Voice communications are available between emergency communications network.
the TSC, Operational Support Center (OSC), and NRC.
The TLorganizatign _has direct access to the main plant _ computer data, base _ that_includ.es_th_e S_afe_ty Parameter Dfspla[Syste]m]fSPDS) whichTonitors and displays information on This enables evaluation of incident sequence, cf1Efcal' safety functions.
appropriate mitigating actions, and damages.
Each TSC will accommodate the technical staff needed to evaluate the plant Communications with and support by the Yankee Engineering Support Center in Framingham, Massachusetts will be an additional aid to the TSC.
condition.
Each TSC will have access to the station Final Safety Analysis Report (FSAR),
the station Emergency Plan and procedures, and a complete set of Station Each TSC will also have access to drawings and equipment specifications.
current records which are essential for evaluation of the Station under accident conditions.
5)
The TSC is habitable to the same degree as the Control' Room for postulated Radiological protection and monitoring equipment assure radiation exposure to any person working in the ISC would not exceed 5 accident conditions.
rem whole body, or its equivalent to any part of the body for the duration of
'that The TSC is designed using good human factors engineering an accident.
principles and is environmentally controlled to provide room air temperature, humidity, cleanliness and lighting approp,riate for personnel and equipment.
The TSC is structurally built in accordance with the Uniform Building Code.
Seis=ic qualifications of the TSC surpass the Uniform Building Code.
i i
Both primary and alternate EOFs satisfy good human factors engineering principles and are environmentally controlled to provide room air temperature, Each EOF humidity, and cleanliness appropriate for personnel and equipment.
is activated to exclude will be provided with industrial security when it unauthorized personnel and when it is idle to maintain its readiness.
Implementation The Seabrook TSC, OSC, and EOF will all be functional prior to fuel load.
ACCIDENT MONITORING INSTRUMENTATION (REGULATORY GUIDE 1.97, REV. 2)
We are in the process of selecting the Seabrook Accident' Monitoring 4.5-1980, " Criteria for Instrumentation ( AMI) using the guidance in ANSI /ANS Accident Monitoring Functions in Light-Water-Cooled Reactors," as endorsed by The plant specific Emergency Operating Regulatory Guide 1.97 (Rev. 2).
procedures identify Procedures (EOPs) and emergency radiological assessmentThe AMI will be selected to the actions required to respond to an accident.
e'nsure that information required to perform these actions is available to
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personnel in the Control Room and the emergency response facilities.
effort involving Engineering and The AMI selection process will be a joint The adequsey of the AMI will be varified as part of the Operations personnel.
Control Room design review.
A list of the Seabrook AMI will be prepared that will provide the information 1.
A comparison to the requested by Section 6.2 of NUREG-0737, Supplement recommendations in Regulatory Guide 1.97 will be provided and exceptions will be Justified.
The AMI list and Regulatory Guide 1.97 comparison will be submitted to the NRC All AMI will be installed, integrated into the E0Ps and by September 1983.
appropriate training perf ormed prior to loading fuel.
EMERGENCY OPERATING PROCEDURES The Westinghouse Owners Group Emergency Response Guidelines (ERGS) are being f
1 used as the basis for the development of the Seabrook Station Emergency Seabrook Station has completed its basic set of Operating Procedures (EOPs).
to the NRC staff's The E0Ps will be revised as necessary subsequent E0Ps.
issuance of its Saf ety Evaluation Report on the ERGS and the Owners Group issuance of Revision 1 to the ERGS.
A review of the E0Ps will be provided by the Seabrook Station Operations Station Operations Review Committee and Westinghouse.
Department management, Verification and validation of the E0Ps will be accomplished on-site, w site-specific simulator.
The E0Ps will be available f or NRC staf f review / audit in December 1983.
Operator training on the E0Ps will begin in March 1984.
L
_a-DETAILED CONTROL ROOM DESIGN REVIEW A detailed Control Room design review of the Main Control Board (MCB) is being lly conducted on the site specific simulator which is maintained as an essentia exact duplicate of the MCB.
A preliminary report on the Control Room design review was submitted to the
("Seabrook Station Control Room Design Review NRC on May 12, 1982.
This submittal provides J. DeVincentis to F. Miraglia. )
0801, and the Preliminary Report";
the preliminary information recommended by NUREG-0700 and NUREG-The preliminary repo 1.
program plan required by NUREG-0737, Supplementindicated that a fin 983.
status is a revised schedule and the current This milestone will not be met-discussed below.
this We have completed a review of all those items that can be checked at There are some These comprise approximately 90% of the full review.is operating before they time.
environmental items which must wait until the plantSpecifically, these are can be checked.
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ventilation.
An assessment of the Human Engineering Discrepancies (HEDs) w identified to date is underway.
from the ongoing assessment of HEDs will be addressed in the Summary Report required by NUREG-0737, Supplement 1, which will be submitted by h
August 31, 1983.-
months subsequent to the date of commercial operation; it will address t e environmental items discussed above.
SAFETY PARAMETER DISPLAY SYSTEM A safety Parameter Display System (SPDS) will be available to Control Room and The status of our design effort on the Technical Support Center personnel.
SPDS is discussed below.
Computer and will consist of a The SPDS will receive its input from the Plant dedicated display (CRT) which monitors the following six critical safety g.
i g) f unc tion:
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Subc ri ticality Core Cooling Reactor Coolant System Integrity Heat Sink Containment Integrity Reactor Coolant System Inventory If the condition of any of the six critical saf ety functions departs from its l
the color of the block corresponding to the effected critica level of saf ety function will vary as follows (see attache-d photograph of this normal range, display):
to critical safety function Green - normal; no threat Yellow - off nor=al; no threat to critical safety function Orange - of f normal; approaching a challenge to the critical safety function
- cff neric1; challenge c the critical safety function Rei
. ih The SPDS can also display a representation of the safety status trees wh c hd
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correspond to any of the six critical safety functions (see attac eThe color of the st photograph of this level of display).
I k (i.e., if the corresponds to the color of the critical safety function blo d if h
will.be green).
the status tree for core cooling is displayed, its branc es d at any The SPDS is capable of displaying the procedure that should ha utilize listings of f
point in the status tree, and is also. capable of providing data various parameters.
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The six critical safety functions and the safety status trees were deve ope from the Westinghouse Owners Group Emergency Response Guidelines h
J It should also be noted that-the Main Control Board will con h SPDS when the analog or. digital instrumentation readouts that are input t d
Therefore, the n
procedures are available to Control Room and TSC pers be manually accomplished if the Plant Computer is unavailable, l}
hics The Plant Computer will use existing software to generate / store the grap Dynamic inputs are fed to the pregenerated/ stored status and the color scheme described above is generated when the input val i
discussed above.
- trees, achieve a preprogrammed level.
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i Some of the safety status trees have already been programmed into the b
i simulator computer and were demonstrated to the ACRS Subcommittea me (Kerr and Michelson) on April 1, 1983.
l simulated SPDS.
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- 1983, Verification and Validation of the SPDS will be accomplished by Decem er at which time operator training will be initiated and a NRCOperators will b post-implementation review can begin.A safety analysis which describes the b 3
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selection is essentially that which has been submitted in the Westing ousetha to fuel load.
It is therefore felt 6
saf ety analysis which provides the basis for SPDS parameter selection.
4 Owners Group ERGS.
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