ML19257A785
| ML19257A785 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/02/1980 |
| From: | Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001080442 | |
| Download: ML19257A785 (12) | |
Text
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POWER AUTHORITY OF THE STATE OF NEW YORK 1C CoLUMaus CIRCLE NEW YORK. N. Y. too19 (2121 397 6200 TRUSTEES GroRGET.eERRY
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FicE cwaiauan RICH ARD M. FLYNN JOSEPH R. SC IEDER ROSERTt. MILLoNCI P4E DENT 4 CMIEF January 2, 1980 rREDERicx R. cLARx t,,oy
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OFFICE 4 Director, Office of Nuclear Reactor Regulation THOM AS R. FREY U.
S. Nuclear Regulatory Commission
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Washington, D. C.
20555 Attention:
Mr. Thomas A.
Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors
Subject:
James A. Fit: Patrick Nuclear Power Plant Docket No. 50-333 Status of "TMI Lesson.s Learned" Related Activities
References:
(1) Letter Harold R. Denton (NRC), to All Operating Nuclear Power Plants dated October 30, 1979 (2) Letter Paul J. Early (PASNY), to Darrell G.
Eisenhut (NRC) dated October 22, 1979 (3) Letter J.
R.
Schmieder (PASNY), to Thomas A.
Ippolito dated November 21, 1979
Dear Siv:
The Authority, by this letter, is providing a status report for activities committed to in Reference 2 and 3 in response to the requirements of Reference 1.
Information has also been pro-vided for NRC implementation review.
Material is attached and ordered by applicable NUREG-0578 requirement number.
The Authority will meet all NRC requested implementation dates except the January 1,1980 date for installation of direct indication of relief valve position (NUREG-0578 requirement 2.1.3.a).
The earliest possible delivery date for the relief valve position indication equipment is March 1980.
The system will be installed within 30 days of receipt.
This schedule has been discussed without objection with NRC staff.
Very truly yours,
.. s..
Paul J.
Early Assistant Chief Engineer-Projects 1706 001 0
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j 2.1.2 Safety Relief Valve Testing The Authority will support the BWR Owners Group ef forts in determining the test conditions and undertake the required testing of the Safety Relief Valve on a generic basis.
Based on the understanding of the Owners Group with Mr. Dave Vearelli of the NRC the Owners Group will be providing a schedule for the testing by 1-31-80.
1706 002 i
Response to 2.1.3.a - Direct Indication of Valve Position The Authority plans to install the Relief Valve Monitoring System furnished by Babcock & Wilcox (B&W).
This system monitors flow in the relief valve discharge line by use of acoustic monitoring instrumentation.
Details of the system design are available upon request.
A Purchase Order for the equipment has been placed, and the earliest possible delivery date for all necessary components is March, 1980.
The delivery schedule has been discussed in detail with the NRC.
Installation will be completed within 30 days of delivery, consistent with discussion with the NRC Project Staff.
B&W is conducting an extensive component qualification program for the acoustic valve monitoring system, which is scheduled for completion by July, 1980.
1706 003
Response to 2.1.4 - Containment Isolation The Authority has completed an extensive review of the FitzPatrick Plant containment isolation systems.
A summary of the review is available upon request.
The results indicate full compliance with the subject position except that Reactor Water Sample Valves 02-AOV-39 and 40 will open after the iso-lation system is manually reset if their control switches are left in the AUTO /OPEN position.
In accordance~with the January 1, 1980 implementation date, the Reactor Water Sample Valve isolation logic has been modified to preclude this condition.
The modification involves connecting a contact from the control switch into the reset circuit of the isolation signal relay.
This modification re-quires the operator to place the control switch in the CLOSE position before reset of the isolating signal.
To open the valve, the operator must then place the control switch in the OPEN position.
1706 004
Response to 2.1.5.a - Dedicated Penetrations for External Recombiner or Post-Accident Purge Systems The Authority has completed an extensive review of the FitzPatrick Plant's ability to meet the captioned NRC position.
A report of the review results is available upon request.
The study indicates that the present design meets the NRC position in all respects except that a potential problem relative to a single failure-proof, remote means to purge containment was indicated.
The Authority is presently inves-tigating this item in detail, and will propose implementation of design changes if necessary on an expedited basis.
1706 005
2.1.6a Post Accident Control of Radiation The FitzPatrick Plant staff has monitored leakage during plant operation from the following systems:
1.
Core Spray (CS) 2.
Low Pressure Coolant Injection (LPCI), including Shutdown Cooling, Containment Cooling, Containment Spray and Vessel Head Spray.
3.
E Pressure Coolant Injec*. ion (HPCI) 4.
W.
Mor Core Isolation Cooling (RCIC) 5.
Scandby Gas Treatment (SBGT)
Zero leakage was detected visibly from Core Spray, LPCI and the Standby Gas Treatment System.
The HPCI and RCIC systems have small steam leaks.
The combined total of the steam leak was less than 0.24 gallons per minute.
The rate of leakage was determined by monitoring the reactor building floor drain sumps and attributing all the input to the HPCI and RCIC leaks.
This number is therefore very conservative.
The HPCI system will be taken out of service in January 1980 to repair its' leaks; after HPCI is restored, the RCIC system leaks will be repaired.
It should be noted that although the LPCI system has no visible leaks during operation, some packing leakage from one of the injection valves was noted during the valve operability test.
This valve does not leak when full open and it is normally open for the system to fulfill its intended function.
It will be repaired during the April, 1980 refucling outage.
The program for reducing leakage is to have the components visually inspected during each of the monthly surveillance tests and any leak reported to the Maintenance Department for repair.
In addition, the s taff has contacted a vendor to determine if helium can be used to determine leak rates in water systems.
It is not certain whether this can be accomplished satisfactorily.
If not, the plant staff will evaluate another method to determine leakage prior to the end of the 1980 refueling outage.
The plant staff is also considering the feasibility of installing timers on each of the reactor building floor drain sumps, to give an alarm should a given amount of input by way of leakage be exceeded.
Modification of the reactor building sump discharge lines is being evaluated.
This modification will re-route the sump discharge to the suppression pool instead of the normal pathway to radwaste.
This alternate path will only be used in post acciden. situations in order to limit radwaste input due to high ictivity leakage in the reactor building.
1706 006
Response to 2.1.6.b - Review of Plant Shielding and Environ-mental Qualification of Equipment Plant Shielding The Authority's Architect-Engineer has completed a compre-hensive review of the FitzPatrick Plant's radiation shielding assuming the conservative post-accident releases of radioactivity specified in the captioned MRC position.
The study concentrated on specific areas of the plant which may require personnel access to mitigate the consequences of an assumed accident.
The radiation study assumptions, inputs and results are available upon request.
A preliminary review of the results by the Authority indi-cates that modifications to reduce radiation levels in several plant areas is necessary.
Review is continuing relative to the specific plant areas, personnel access times required in these areas, and the means to reduce radiation levels; i.e.,
additional shielding, modification of procedures, etc.
Subsequent to completion of this review, our Architect-Engineer will be directed to proceed with detailed design and installation of the modifications necessary on a best effort basis.
Equipment Qualificaticn A study.was conducted by the Authority's Architect-Engineer relative to the qualification of components in systems which could be used to mitigate the consequences of an accident, and could carry highly radioactive fluid.
The fundamental purpose of the study was to determine which components were qualified for exposure to highly radioactive fluids over a long period of time.
The study's assumptions, inputs, and preliminary results are available upon request.
The principal conclusion reached is that all system equipment will perform their safety function throughout the accident, but that components utilizing Teflon packing or gasket material would require additional qualifica-tion or change-out to a material more resistant to high levels of radiation.
The Authority, in conjunction with our Architect-Engineer, is continuing the review of equipment qualification on an expedited basis to determine the full extent of additional-qualification or equipment change-out which may be required.
1706 007
3 Response to 2.1.8.a - Post Accident Sampling A design review has been conductec hy the Authority's AE and recommendations for plant modificacions have been made to upgrade the post accident sampling and analysis capability.
The proposed modifications are currently being reviewed by the Authority's staff.
Conceptual system descriptions have been prepared and are available upon request.
Subsequent to review, our Architect-Engineer will be directed to complete the detailed design, procurement and installation functions as necessary, and on an expedited basis.
In the interim the Authority has developed a sampling and analysis procedure for high activity reactor water samples.
This procedure has been reviewed by the Plant Operations Review Committee (PORC).
approved for use, and is available for inspection at the plant.
1706 008
.m 2.1.8b Increased Range of Radiation Monitors The Autnority has developed a specification for a high range monitoring system.
The monitoring system will include tne capability of analyzing noble gas and iodine effluents from the various plant release points as well as in-containment radiation levels.
Solicitation for bids on the monitoring system will be made in the near future and the system is expected to be operational by January 1, 1981.
The Authority has developed interim procedures for monitoring noble gas effluents in a post accident situation.
These procedures have been review by PORC and approved for use.
Noble gas release rates are determined by using portable high range ion chambers mounted in a marinelli sampling chamber.
Samples are drawn from existing isokenetic probes and the flowing sample is analyzed in a low background area.
This procedure is capable of determining the concentration of noble. gases over a range of 10-1 to 10+4 micro Curie /CC.
Ccmmunications between the " sampler" and the control room (and/or Emergency Director) are maintained using Portable FM Radios.
1706 009
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2.1.8.c Improved In-Plant Iodine Instrumentation Under Accident Conditions The Authority has reviewed the existing plant procedure for analysis of iodine concentracions may be present during an accident and finds chat they are adequate to meet the NUREG-0578 position.
1706 010
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2.1.9 Transient and Accident Analysis The Authority has developed new procedures and had comp-leted the training of the plant operators based on the guidelines worked out by the BWR Owners Group and appro-ved by the B&O Task Force of the NRC.
Development of future procedures and associated training will be compl-eted after guidelines are developed by the Owners Group and af ter review and approval by the NRC per the schedule agreement reached by the Owners Group with the NRC.
It is the Authority's understanding that the General Electric Company has been in direct contact with the NRC on the Two Loop Test Apparatus (TLTA) preprediction and Small Break Test in support of Transient and Accident Analysis Effort.
As per General Electric, the direct submittal will be made to the NRC by December 28, 1979.
Containment Instrumentation The Authority has conducted a review of the FitzPatrick Plant containment pressure, hydrogen and water level instru-mentation relative to the requirements of the captioned NRC position.
The review indicates that some additional instrumentation will be required.
The Authority is pursuing procurement of the necessary instrumentation on a best effort basis.
1706 011
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24 2.2.la Shift Supervisors Responsibilities The Authority has up-dated procedures on the Shift Super-visors Responsi )ilities which have been reviewed and ap-proved by th'e Einior Vice President and Director of Power OperationD.
2.1.lb Shift Technical Advisors The Authority will have Shift Technical Advisor's on shift by 1-1-80.
2.2.lc Shift Turnover Procedures The subject procedures have been up-dated, approved and implemented.
2 ~ 2a Control Room Access Control Procedures for access control have been reviewed, Approved and implemetted.
2.2.2b On-Site Technical Support Center The Authority has established the On-Site Technical Support Center at Fit: Patrick plant.
The center contains control set of plant drawings, tech-nical manuals u.c modification packages.
A tie-in wir.h the plant computer has been established repeating the control room video parameters and alarm displays.
An additional plant computer tie-in is also provided to permit line printer display of about 200 plant parameters.
This additional feature is indepe-ndant to the control room and easily expanded up to 640 parameters.
The Technical Support Center conta2ns the following communication facilities.
1.
Radio Base Station 2.
NRC Emergency Red Telephone 3.
Dedicated Control Room Phone Jack 4.
Ten Telephone Extensions 5.
Two Plant Paging Stations The Authority is evaluating a modification to pro-vide filtered ventilation for the center in the near future.
2.2.2c On-Site Operational Support Center The Authority has an On-Site Operational Support Center established in the corridor adjacent to the control room.
1706 012
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