ML20204A281
| ML20204A281 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/05/1986 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| References | |
| GL-84-09, GL-84-9, TAC-58018, TAC-59829, TAC-60152, TAC-60153, NUDOCS 8605120020 | |
| Download: ML20204A281 (7) | |
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NUCLEAR REGULATORY COMMISSION
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May 5, 1986 Docket No. 50-219 Mr. P. B. Fiedler Vice President and Director Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Fiedler:
SUBJECT:
MEETING OF APRIL 10, 1986, ON REQUESTED CANCELLATION OF NITROGEN PURGE / VENT SYSTEM (TAC 59829)
Re:
Oyster Creek Nuclear Generating Station
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In the neeting of April 10, 1986, the staff and GPU Nuchar personnel (the licensee) discussed the following Oyster Creek licensing actions before the staff: (1) cancellation of two torus pool modificati(ns, (2) Generic letter 84-09 dated May 8, 1984, and (3) cancellation of upgradina the These are licensing actions associated with nitrogen purae/ vent system.
the Cycle 11 Refueling (Cycle 11R) outage which began April II, 1986, and is scheduled to end in October 1986. The expectation was that these actions would be resolved before the commencement of the current outace.
The staff requested information on the torus pool modifications and Generic Letter 84-09, in the April 10, 1986, meeting in order to complete its evaluation. The requested information is an enclosure to this letter.
Two important conclusions were drawn at the April 10, 1986 meeting:
(1) the control on oxygen in the inerted containment at Oyster Creek during the design basis Loss-of-Coolant Accident (LOCA) is not sufficient in itself to be the safety-g)rade combustible gas control system (CGCS) required per 10 CFR 50.44(g (2) the initial suppression pool temperature used in the LOCA analysis may be less than the pool temperature limits in the Oyster Creek Technical Specifications (TS) and may, for specific transients, result in loss of net A
positive suction head (NPSH) for the core spray pumps during a LOCA.
question on item (2) is also in the enclosure to this letter.
The staff's position is that RG 1.7 is predicated on defense-in-depth and, because of LOCA combustible oas uncertainties, is the appropriate source terms to be used in LOCA analyses. With your inerted containment at less than or equal 4% oxygen and following the guidelines of RG 1.7, the oxygen I
concentration in the containment will be above 5% oxygen within 30 days of a It is also the staff's position that a safety grade CGCS design basis LOCA.
in addition to an inerted containment is required to meet the requirements i
of 10 CFR 50.44(g).
It is our understanding that you believe RG 1.7 is 8605120020 860505 I
PDR ADOCK 05000219 i
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9 overly conservative and, using the source terms of NED0-22155, the inerted containment is sufficient in itself to meet 10 CFR Part 50.44(q). The source terms in this NED0 report have been reviewed by the staff. The staff ccncluded in GL 84-09 that, relative to the need for recombiner capability, the NEDO report provided a basis to accept the RWR Mark I Owner's Group position that Mark I plants do not rely on the use of the safety orade purce/
repressurization as the primary means of combustible gas control. The staff concluded that both the inerted containment and the purge /repressurization were the means of combustible gas control. The staff in this review, however, still concluded that R.G. 1.7 was appropriate for the LOCA analysis to meet 10 CFR 50.44(g).
We require that you commit to provide a safety-grade CGCS (e.g., a containment purge and/or repressurization system) by the restart from the Cycle l?R outage to meet 10 CFR 50.44(g).
It is also the staff's position that the initial suppression pool temperature for a LOCA shall be consistent with the most conservative pool temperature allowed by the TS for any extended period of time. Using this analysis there should not be NPSH problems with the operation of a safety-grade system such as the core spray system during a design basis LOCA. The core spray system is relied upon in accident analyses to protect the core during a LOCA. This staff position was discussed with the licensee during the April 10, 1986, meeting and by telephone on April 11, 1986.
You should also address the questions enclosed in this letter.
For the question on the initial suppression pool temperature during a design basis LOCA, you should provide a schedule for completing any calculations needed to answer the ouestion. You should also provide any compensatory measures, if needed, to account for the LOCA calculations not being based on the appropriate suppression pool limits in the TS. This may reouire new TS limits based on the LOCA analyses. These limits would be required prior to the restart from the Cycle 11R outage.
Because these issues are important and are now clearly focused, the staff desires to complete these technical issues as soon as reasonably possible.
We request your response to this letter addressing the above issues by May 15, 1986.
The reporting and/or recordkeepino requirements contained in this letter affect fewer than ten respondents; therefore, OMR clearance is not required under P.L.96-511.
Sincerely, ORIGI ML SIG ED BY John A. Zwolinski, Director RWR Pro.iect Directorate #1 Division of BWR Licensino
Enclosures:
Requests for additional DISTRIBUTION information Docket File EJordan GHolahan NRC PDR BGrimes Glainas cc w/ enclosures:
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j overly conservative and, using the source terms of NED0-?2155 t containment is sufficient in itself to meet 10 CFR Part 50.44fq)he inerted The /
source terms in this NE00 report have been reviewed by the staff. Thep taff concluded in GL 84-09 that, relative to the need for recombiner capa)111ty, the NED0 report provided a basis to accept the BWR Mark I Owner's fp'oup position that Mark I plants do not rely on the use of the safety rade purce/
repressurization as the primary means of combustible gas control The staff concluded that both the inerted containment and the purge /reer surization were the primary means of combustible gas control. The staff in this review, however, still concluded that R.G. 1.7 was appropriate for he LOCA analysis to meet 10 CFR 50.44(g).
We request that you comit to provide a safety-grade Cr S (e.g., a containment ourge and/or repressurization system) by the restart # om the Cycle 12R outage to meet 10 CFR 50.44(g).
It is also the staff's position that the initial ppression pool temperature for a LOCA shall be consistent with the most con ervative pool temperature allowed by the TS for any extended period of t
.e. Using this analysis there should not be NPSH problems with the op ation of a safety-grade system such as the core spray system during p design basis LOCA. The core spray system is relied upon in accident anafyses to protect the core during j
a LOCA. This staff position was discusse with the licensee during the April 10,1986, meeting and by telephone on April 11, 1986.
You should also address the questions nclosed in this letter.
For the question on the initial suppression 001 temperature during a design basis LOCA, you should provide a schedule for completing any calculations needed to answer the question. You shoul also provide any compensatory measures, if needed, to account for the LO calculations not being based on the appropriate suppression pool li its in the TS. This may require new TS limits based on the LOCA analy es. These limits would be required prior to the restart from the Cycle 1 outage.
Because these issues are i portant and are now clearly focused, the staff desires to complete these technical issues as soon as reasonably possible.
We request your response to this letter addressing the above issues by May 15, 1986.
The reporting and/or ecordkeeping requirements contained in this letter affect fewer than t respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing Enclosures Requestsforadditional DISTRIBUTION information Docket File EJordan GHolahan cc w/,e/
NRC PDR RGrimes Glainas nclosures:
Local PDR JPartlow GHulman See next page BWD#1 Reading CJamerson JKudrick j
OELD JDonohew ACRS (10)
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Mr. P. B. Fiedler Oyster Creek Nuclear Oyster Creek Nuclear Generating Station Generating Station CC:
Ernest L. Blake, Jr.
Resident inspector Shaw, Pittman, Potts and Trowbridge c/o U.S. NRC Post Office Box 445 1800 M Street, N.W.
Washington, D.C.
20036 Forked River, New Jersey 08731 Commissioner J.8. Liberman, Esoufre Bishop, Liberman, Cook, et al.
New Jersey Department of Energy 1155 Avenue of the Americas 101 Commerce Street New York, New York 10036 Newark, New Jersey 07102 Eugene Fisher, Assistant Director Regional Administrator, Region I Division of Environmental Quality U.S. Nuclear Deaulatory Commission Department of Environmental Protection 631 Park Avenue King of Prussia, Pennsylvania 19406 380 Scotch Road Trenton, New Jersev 08628 RWR Licensino Manager GPU Nuclear 100 Interoace Parkway Parsippany, New Jersev 07054 Deputy Attorney r,eneral State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 West Lacey Road Forked River, New Jersey 08731 D. G. Holland Licensing Manager Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
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REQUEST FOR ADDITIONAL INFORMATION OYSTER CREEK NUCLEAR PLANT DOCKET NO. 50-219; TAC NO. 58018 By a letter dated August 14, 1985 the licensee responded to the staff's request for additional information (RAI) dated April 29, 1985.
The following RAI is based on the licensee's August 14 submittal:
In the response to question 1, you stated that the backup air 1.
supply to the Nitrogen System will be automatically isolated when the primary containment isolation occurs for the design basis Loss-of-Coolant Accident (LOCA).
Identify the components relied upon to isolate the backup air supply from the Nitrogen System and verify that they are Also, indicate if the response is valid for the situation
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safety grade.
of a failed-nitrogen system. Describe the inspection and testing progam employed to assure the operability of these components.
In the response to question 2 you stated that the service air 2.
and breathing air systems are not connected to the drywell during power operation.
Furthermore, you stated the TIP purge system which may use nitrogen or air, uses nitrogen during power operation.
Describe the administrative controls and/or interlocks used to prevent these systems from adding oxygen to the containment during power operation.
Identify the components relied upon to isolate these potential oxygen contributors from the containment, and verify that these components are safety grade.
Describe the inspection and testing program employed to assure the reliability of these components.
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RE00EST F00 ADDITIONAL INFORMATION OYSTER CREEK NUCLEAR PLANT DOCKET NO. 50-219; TAC NO. 60157, 60153 By letter dated October 31, 1985, the licensee reouested cancellation of two modifications that were to be completed in the torus suporession pool The following information is reeded by the during the Cycle 11 R outage.
staff to complete its evaluation:
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Is the Acceptance Criteria on safety relief valve discharges in 1.
Appendix A to NUREG-0661, Mark I Containment Long-Term Procram, applicable to these modifications and does Oyster C eek neet the This includes the use Acceptance Criteria without the modifications.
of the Monticello test data to obtain the 43*F as the maximum Pas local-to-bulk torus water temperature difference for Ovster Creek.
anything changed such that Oyster Creek is outside the assumptions made by the staff's two safety evaluations dated January 13, 1984, on the Mark i Containment Long Tem Program.
In the April 10, 1986, meeting on the above technical issues, the licensee stated that the initial suppression pool temperature used in the LOCA may be less than the most conservative pool temperature allowed ic the Technical 3
For the staff to under-Specification (TS) for any extended period of time.
stand what should be the appropriate TS on the suporession pool temperature, you should provide the following information requested in the meeting:
Discuss the highest torus suppression pool temoerature assumed in the 1.
Loss-of-Coolant Accident (LOCA) analysis and net positive suction head The statements made by the licensee l
(NPSH) of the core spray pumps.
are that these pumos would be the first safety pumps which would Was the initial f
experilnce loss of NPSH as the pool temperature rises.
l LOCA pool temperature based on the suppression pool temperature limits in the station Technical Specifications (TS). The initial temperature should be the most conservative limits allowed by tha TS for any
,j extended period of time, i
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,. Although the followina detailed information was not reouested in the April 10, 1986, meetina, your discussion on suopression pool temperature and NPSH for the core spray pumps s'nould include the following:
Describe the core spray pump suction header lecations in the suppres-2.
sion pool in relation to the relief valve (RV) quenchers, to the suppression pool bulk temperature monitors, and the pool water level, Discuss the representativeness of the suppression pool terperature 3.
monitors for measuring the temperature at the core spray suction in the pool during RV blowdown, Discuss any assumptions made in the calculation of the temperature at 4
core spray suction in the pool that take credit for pool mixino during RV blowdown to the torus, and What is the maximum temperature at the core spray suction header 5.
during RV blowdown in the LOCA, and what are your bases?
Compare your assumptions in your calculations for NPSH to the f
6.
Regulatory Position in Regulator _y Guide 1.1, Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps,"
dated November 1970, i
References to your Updated FSAR'as part of your response are acceptable.
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