ML19210E284
| ML19210E284 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/20/1979 |
| From: | Pilant J NEBRASKA PUBLIC POWER DISTRICT |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7912040251 | |
| Download: ML19210E284 (8) | |
Text
e' GENERAL OFFICE P. O. BOX 499, COLUMBUS, NEB R ASKA 68601 A vb. Nebraska Publ.ic Power Distr. t TETE ~o~E <4ou se4 8sei
-e ic November 20, 1979 Mr. Darrell G. Eisenhut, Acting Director Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Lessons Learned Short Term Requirements Committal Cooper Nuclear Station NRC Docket No. 50-298, DPR-46
Reference:
- 1) Letter from J. M. Pilant (NPPD) to D. G. Eisenhut (NRC)
Dated October 17, 1979, "Three Mile Island Follow-up Actions"
- 2) Letter from D. G. Eisenhut (NRC) to All Operating Nuclear Power Plants, dated September 13, 1979,
" Follow-up Actions Resulting from the NRC Staff Resiew Regardi/3 the Three Mile Island Unit 2 Accident"
- 3) Letter from H. R. Denton (NRC) to All Operating Nuclear Power Plants, dated October 30, 1979,
" Discussion of Lessons Learned Short Term Requirements"
- 4) Letter from T. D. Keenan (Vermont Yankee - BWR Owners Grou i) to D. G. Eisenhut (NRC), dated October 17; 1979, "BWR Owners Group Positions On NUREG 0578" Dei.. Mr. Eisenhut:
Per t' a commitment made in Reference 1, attached are Nebraska Public Power District's commitments to meet all of the requirements specified in Reference 2 and clarified in Reference 3.
As required in Reference 3, a revisic a of the schedule for Staff Position 2.1.3.a " Direct Indication of Valve E asition" is requested. A justifi-cation for the delay and a description of our degree of compliance by January 1, 1980 is included. All othcr Short Term Requirements will be complied with per the NRC schedule.
hyo 1470 240 s
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Mr. Darrell G. Eisenhut November 20, 1979 Page 2 If additional clarification is necessary regarding any of the enclosed information, please do not hesitate to contact me.
Sincerely, M
Jay i. Pilant Director of Licensing and Quality Assurance JD'l/cmk Attachment 1470 241
Nebraska Public Power District Commitments to Requirements of NUREG 0578 NUREG 0578 Section 2.1.1 There is no need for action in response to this requirement for the reasons stated in Emergency Power Reference 4.
Supply Requirement 2.1.2 As concluded in Reference 4, concerns regarding safety / relief valve performance have Relief and Safety been addressed and no special performance testing is considered to be required.
In Valve Testing the absence of formal NRC comment on the implementation criteria submitted in Reference 4, NPPD will meet the BWR Owners Group implementation criteria. However, if the NRC does not fully accept the BWR Owners Group criteria, more discussion between the BWR Owners Group and the Staf f will be required to further clarify valve testing require-ments af ter which, NPPD will embark upon the required testing program in conjunction with the Owners Group.
2.1.3.a Pressure switches will be installed which meet the NUREG 0578 position and the BWR Direct Indication Owners Group implementation criteria. General Electric Co. has been contacted to of Valve Position provide the necessary hardware; however, delivery of the printed circuit boards necessary for the processing of multiple inputs and the interf acing with multiple outputs will reovire at least 12 weeks.
The Cooper Nuclear Station refueling outage will begin on or about March 15, 1980.
During this outage major modifications to the safety / relief valves, safety / relief valve discharge piping, and torus will be completed.
Installation of a means for a direct positive indication of relief valve positions will be implemented into the total design of the pressure suppression system at that time.
It should be noted that the necessary environmental qualification may not be available prior to instal-lation.
Cooper Nuclear Station currently has a method and procedures to deternine if a safety /
relief valve has opened and has not closed. A thermocouple is installed on the relief valve discharge tail pipe. Operability and response of this thermocouple is tested by actuating the relief valve at 150 psi pressure during each startup after relief valve maintenance has been performed. These relief valve discharge temperatures are con-
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tinuously recorded on a recorder.
An alarm is actuated when the temperature exceeds setpoint. The process computer also prints out on the events log when a discharge C-)
temperature reaches alarm setting. Although the temperature monitoring method does ng have a slow response in indicating valve closure, this slowness does not influence
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further operator actions as the operator must be most concerned about pressure and rs) water level which can be quickly monitored.
NUREG 0578 Section 2.1.3.a Simultaneous to, and following relief valve opening, other events occur that indicate a Continued relief valve has opened. The suppression chamber water and air temperature increases, the suppression chamber pressure increases and, of course, the reactor pressure decreases.
All of these events are recorded. The relief valve actuation also causes momentary actu-ation of suppression chamber level lif /Lo alarm. A rumbling noise in the reactor building is easily distinguishable for additional confirmation.
Actuation of the relief valves and the subsequent above indications are expected events following certain scr'am and isolation events. Thus, data are available and operators 3
are trained to understand these events.
If a valve does not close, it would be readily z
detectable because of the continuation of the above indications. Knowing that a relief valve is stuck open or which relief valve is stuck open, is of minimal value to the operator since he has no means to close the valve. Ile can only reactuate the valve which, from past experience, does not assure that the valve will close. The operator must be prepared for the valve to remain in its open position and handle the subsequent rapid depressurization.
We believe that the above provisions are adequate to justify continued operation until the scheduled refueling outage at which time a more direct positive indication of valve position will be installed.
2.1.3.b NPPD will meet this position as described in the BWR Owners Group Implementation Instrumentation for Criteria.
Inadequate Core Cooling 2.1.4 NPPD will meet this position as described in the BWR Owners Group Implementation Diverse Containment Criteria.
Isolation 2.1.5.a This pos' tion is not applicable to Cooper Nuclear Station since the licensing basis Dedicated Ilydrogen did not include requirements for external recombiners or purge systems for post-Control Penetrations accident combustible gas control of the containment atmosphere.
2.1.5.c This position is not applicable to Cooper Nuclear Station since the licensing basis 42-Recombiner did not include requirements for hydrogen recombiners.
'sJ Procedures CD 2.1.6.a Nebraska Public Power District will meet this NRC position as described in the following I'5 Systems Integrity BWR Owners Group position:
[ for liigh Radio-activity
NUREC 0578 Section 2.1.6.a Practical leakage reduction measures will be investigated for systems which may contain Continued radioactive fluids post-LOCA.
Such systems as the reactor core isolation cooling system, high-pressure coolant injection system, core spray system, residual heat removal system, and waste disposal system will be examined.
This examination will include a study of valve stem packing leakoffs, rotating seals on equipment, gasketed connections or joints, drains piped to open connections, and reactor drainage system.
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Those components in the above systems from which leakage may be measured will be identified and measured leakage f rom these components will be reported to the NRC.
A periodic leak inspection program will be impicmented on these components.
The above investigations will be completed by January 1,1980.
2.1.6.b Nebraska Public Power District will meet this NRC position as described in the following Plant Shieldit '
BWR Owners Group position:
Review BWR plants are specifically designed to mitigate major design basis events with no access outside the main control room being required. With this goal in mind, the plants were not specifically designed for any access outside the main control room.
To specifically design for guaranteed access at anytime in most parts of the reactor building is not feasible. Iloweve r, the current designs may allow for access for short times if the entry time into the area can be selectively chosen. Design changes in shielding will be made if evaluations identify feasible modifications which should significantly enhance desirable access. The guidelines for the evaluations are given below.
A TID 14844 radioactivity release will be assumed into the primary containment. A summation of the radioactivity levels from sump water leakage from process systems in the reactor building will be made. The next step will be to calculate the source terms for the suppression pool recirculating piping, pumps, and valves installed in the reactor building assuming that a TID 14844 release had occurred. The vital areas will be identified in the reactor building which may need to be entered during an accident recovery period. The shielding in these vital areas will be reevaluated to 4,
assess its effectiveness in such a circumstance. The occupancy time limits, taking
'*J into consideration transit time, airborne radioactivity levels, and gamma shfne CD intensities, will then be calculated for the vital reactor building areas.
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The above shielding review will be completed by January 1,1980.
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NUREG 0578 Section 2.1.7.a This position is not applicable to Boiling Water Reactors.
Auto Initiation of Auxiliary Feed 2.1.7.h This position is not applicable to Boiling Water Reactors.
Auxiliary Feed Flow Indication I
2.1.8.a Nebraska Public Power District will meet this NRC position as described in the following j
Post-Accident BWR Owners Group position:
Sampling A design and operational review of existing reactor coolant and containment atmosphere sampling f acilities will be completed by January 1,1980.
Procedures have been devised to evaluate the primary coolant system and containment environment activity depending on the accessibility of the sampling stations for particular degraded conditions. The procedures are presently in the final review process and will be in place by January 1, 1980.
Modifications will be made to provide the capability to promptly obtain pressurized and unpressurized reactor coolant samples and containment atmosphere samples.
Analysis capability shall be provided to identify and quantify (1) certain isotopes that are indicators of core damage (i.e., noble gasca, iodines and cesiums, and non-volatile isotopes), and (2) dissolved gases (i.e.,112 and 0 ) and boron concentration 2
of liquils. These modifications will be complete by January 1, 1981.
2.1.8.b NPPD will implement the requirements of position 2.1.8.b items 1, 2, and 3 as clarified liigh Range in Reference 3.
A purchase order has been issued to General Electric Co. to perform Radiation Monitors generic prototype equipment design and qualificetion services.
Procedures will be in place by January 1,1980 to u.termine (or estimate) noble gas and radioiodine release rates if the existing effluent instrumentation goes off scale.
2.1.8.c NPPD will implement this requirement consistent with the guidance of Reference 3 and Improved Iodine develop procedures to accurately determine in-plant lodine concentrations by Instrumentation January 1, 1980.
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%J 2.1.9 The specific requirements and schedules relating to this position are being developed CD Transient and in a continuing series of meetings between the BWR Owners Group and the NRC Bulletins Accident Analysis and Orders Task Force.
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NUREG 0578 Section 2.1.9 The implementation of emergency procedures and retraining will be done on a schedule Continued consistent with those established with the Bulletins and Orders Task Force.
2.2.1.a This position as clarified in Reference 3 will be implemented by January 1,1980 Shif t Supervisor subject to the following interpretation:
Responsibilities In order to remove any ambiguity from the meaning of the term "recident situation" in item 2.b of the Staff's position in Appendix A of NUREG 0578, the entire sentence will be interpreted as "The shif t supervisor, until properly relieved, shall remain l
in the control room at all times whenever a general emergency has been declared to direct the activities of the control room operators."
2.2.1.b Nebraska Public Power District fully agrees with the BWR Owners Group position on Shift Technical this issue presented in Reference 4.
Based on the operating experience gained at Advisor Cooper Nuclear Station, it must be reemphasized that a Shift Technical Advisor offering an independent assessment of an accident to the Shift Supervisor would result in confussica, conflict, and delay in the response necessary during the immediate phase of accident assessment.
As pointed out in the ACRS letter to the Chairman, USNRC, dated August 13, 1979, manpower requirements for the STA objective necessitate that alternate solutions be considered.
Commencing January 1, 1980, NPFD will assign an additional Senior Reactor Operator to each shift at Cooper Nuclear Station who will meet the objectives of this recommendation. Tests conducted at CNS have assured that the necessary qualified personnel required to meet the NRC objectives during the Intermediate Phase of accident assessment will be on-site within 30 minutes of being notified by the Shift Supervisor.
2.2.1.c This position will be implemented by January 1,1980 as follows:
Shift and Relief Turnover Procedures
- 1) A checklist will be devised to ensure that control room status of systems that are required to mitigate the consequences of an accident are monitored on a shift turn-over basis. This list will include system lineups and alarms located in the main control room.
Systems and components in a degraded condition will be identified as required by plant status.
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- 2) The checklist will be kept in the control room at all times.
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- 3) The checklist will be reviewed by personnel other than the shift supervisor as appropriate.
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NUREG 0578 Section 2.2.2.a Procedures will be developed and implemented by January 1, 1980 which meet this position.
Control Room Access Control 2.2.2.b An Onsite Technical Support Center in the Computer Room adjacent to the Control Room Onsite Technical will be established by January 1,1980 which conforms to the BWR Owners Group Imple-Support Center mentation Crtieria presented in Reference 4.
NRC habitability and emergency power requirements will be met.
The clarifications provided by Reference 3 will be incor-
- porated, j
2.2.2.c An Onsite Operational Support Center will be established by January 1,1980 as defined d
Onsite Operational in Reference 4.
Support Center N
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