ML20003F179
| ML20003F179 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/09/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-15-11, TASK-15-13, TASK-RR LSO5-81-04-011, LSO5-81-4-11, NUDOCS 8104200361 | |
| Download: ML20003F179 (6) | |
Text
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e April 9, 1981 Docket No. 50-219
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9! nPp Mr. I. R. Finfrock, Jr.
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Vice President - Jersey Central WlA
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Power & Light Company Post Office Box 388 M'
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Forked River, New Jersey 08731
Dear Mr. Finfrock:
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SUBJECT:
OYSTER CREEK - INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION (SEP TOPIC XV-ll) AND l
SPECTRUM OF ROD DROP ACCIDENTS (SEP TOPTC XV-13)
Enclosed are topic evaluations for SEP Topics XV-11 and XV-13. This evaluation for XV-13 does not include assessment of the radiological l
consequences, which will be issued by the staff at a later date. These eva'uations compare your facility with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment teithin 30 days of receipt of this letter.
These evaluations will be basic inputs to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the l
as-built conditions at your facility. These topic assessments may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crurschfield, Chief Operating Reactors Branch No. 45 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
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% *.....o' April 9, 1981 Docket No. 50-219 LS05-81-04-Oll Mr. I. R. Finfrock, Jr.
Vice President - Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
0YSTER CREEK - INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION (SEP TOPIC XV-ll) AND SPECTRUM OF R00 DROP ACCIDENTS (SEP TOPIC XV-13)
Enclosed are topic evaluations for SEP Topics XV-ll and XV-13. This evaluation for XV-13 does not include assessment of the radiological consequences, which will be issued by the staff at a later date. These evaluations compare your facility with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.
These evaluations will be basic inputs to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. These topic assessments may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated 1
assessment is completed.
Sincerely, Dennis M. Crutchfield, Ch' Operating Reactors Branc
- 40. 5 l
Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page
Mr. I. R. Fi nf rock, J r.
CC G. F. Trowbridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiaticn Protecticn Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 0862S GPU Service Corporation ATTN: Mr. E. G. Wallace Commissioner Licensing Manager New Jersey Department of Energy 260 Cherry Hill Road 101 Cornerce Street Parsippany, New Jersey 07054 Newark, New Jersey 07102 Natural Resources Defense Council Plant Superintendent 91715th Street, N. W.
Oyster Creek Nuclear Generating Washington, D. C.
20006 Station P. O. Box 388 Forked River, New Jersey 08731 Steven P. Russe, Esquire 248 Washington Street Resident Inspector P. O. Sex 1C60 c/o U. 5. NRC Tc s River, New Jersey 05753 P. O. Box 445 Forked River, New Jersey 08731 Jcseph W. Ferraro, Jr., Esquire Deputy Attcrney General Director, Criteria and Standards State of New Jersey Division Department cf Law and Public Safety Office cf Radiation Prograts 1100 Ray:cnd Beulevard (ANR-460)
Newark, New Jersey 07012 U. S. Environmental Protection Agency Ocean County Library Washington. D. C.
20450 Brick Townsni; Branen 401 Chancers Bridge Reac U. S. Envirennental Prctection Brick Town, New Jersey 0S723 Agency Regien II Office Mayor ATTN: EIS C00RDINATCR Lacey Township 25 Federal Plaza P. O. Box 475 New Y crk, New Y crk 10007 Ferked River, New Jersey CS731 Cc:missioner Department of Public Utilities State of New Jersey 101 Correrce Street Newark, New Jersey 07102 i
FUEL MISLCADING EVENT AT OYSTER CREEK (XV-11)
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f The fuel misloading event censists of tne inadvertent loading and operation l
of a fuel assenbly in an improper position. Two different events are con-4 sidered - a fuel assembly loaded into an improper core 1ccation, and an assembly loaded into tne precer location but improperly criented (i.e.,
i rotated by 90 cr 150 degrees).
The consequences'of these events have been analy:ed in spite of the icw probability of their occurrence. The acceptance criterion for this event (Standard Review Plan, Section 15.4.7) is that ner=al cperation with a misicaded assecoly that cannot be detected by plant instrumentaticn shall not lead to violation of fuel thermal limits (linear heat generation rate or minimum critical pcwer ratio).
The fuel misleading event is treated as part of tne design precedure for each reload. The procedure to be emoloyed in tne future is described in NEDO-24195, " General Electric Reload Fuel Application for Oyster Creek."
This precedure has been reviewed and approved by tne staff for use in boiling water reactors.
Its use is acceptable for Oyster Creek.
The procedure in use at present (with fuel supplied by ancther vendor) nas been approved previously (see in particular the evaluation for Amendment 9 to the Oyster Creek Technical Specifications dated May 24,1975). Tne acceptance criteria for this event are similar to those in use at other BWR's.
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! We conclude, based on our review, that, with respect to the ~ analysis of the
' Fuel Misloading Event the Oyster Creek reactor meets the criteria which are l
applied to present generation boiling water reactors, and is therefore acceptable, f
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ROD OROP ACCIDENT IN OYSTER CREEK (XV-13) f The rod drop accident in a boiling water reactor occurs when a rod blade becomes disconnected from its drive and is stuck in the upper portion of the core. The drive is then withdrawn leaving the rod behind. At scme later time, the rod blade becomes unstuck and falls -apidly cut of tne core.
If the control rod has sufficient worth the potential for localized fuel damage exists.
In order to limit the worth of a potential orooped rod, a rod withdrawal sequence is defined and enforced, in Oyster Creek, by a Rod Worth Minimizer, a computer based device which blocks rod motion if an attempt is made to withdraw an out of sequence rod. The analysis of the rod drop event is then done for the maximum potential dropped rod und2r the assumption that the sequence is followed.
The analysis of the rod drop accident is made generically. That is, curves of peak fuel enthalpy as a function of dropped rod worth are drawn for what are believed to be conservative values for certain parameters.
These parameters include the Doppler coefficient of reactivity, the shape of the scram curve, the rod drop speed and the scram insertion i
time. For each cycle the input parameters are compared to those in the generic analysis and if all are conservative then the generic analysis is accepted.
If any parameter is nonconservative, then a cycle specific analysis is performed or use is made of sensitivity studies performed around the base case generic analysis.
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The base case generic analysis which has, up to now, been employed at Osyter Creek is described in Amendment 74 to the Oyster Creek FDSAR
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submitted in a letter (I. R. Finfrock (JCPL) to K. R. Goller (USAEC) h dated May 31,1974). This submittal also describes the procedures to l
the used on a cycle-to-cycle basis. The generic analysis to be used in i
the future is described in fled 0-10527, " Rod Drop Accident Analysis for Large Beiling Water Reactors," and in Supplements I and 2 to that report.
The cycle-to-cycle procedure is described in ilE00 24195, " General Electric Reload Fuel Application for Oyster Creek." The two procedures are essentially the same. The acceptance criterion in both analyses for fuel enthalpy is 280 calories per gram radially averaged enthalpy in the peak pellet.
C'n the bases that the analyses procedures have been and continue to be acceptable and that the fuel enthalpy consecuences meet our acceptance criterion of 280 calories per gram, we conclude that, with respect to the spectrum of rod drop accidents the Oyster Creek analysis meets current criteria and is acceptable.
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