ML20003E681
| ML20003E681 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/31/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-15-08, TASK-15-8, TASK-RR LSO5-81-03-083, LSO5-81-3-83, NUDOCS 8104080373 | |
| Download: ML20003E681 (6) | |
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o UNITED STATES
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NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555 g
March 31, 1981 Docket No. 50-219
{ Mm LS05-81-03-083 so 4
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% pY Mr. I. R. Finfrock, Jr.
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Op/S9g, Vice President - Jersey Central
-11 Power & '.ight Company M
y Post Office Box 388 p
Forked River, New Jersey 08731 g
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Dear Mr. Finfrock:
SUBJECT:
OYSTER CREEK - SEP TOPIC XV-8, CONTROL R0D MISOPERATION Enclosed is a copy of our evaluation of SEP Topic XV-8, " Control Rod Misoperation." This assessment compares your facility with the criteria currently used by the regulatory staff for licensing new facilities.
Please inform us if your facility differs from the licensing basis assumed in this assessment within 30 days of receipt of this letter.
This topic evaluation will be a basic input to the integrated safety assessment unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is complete.
l Sincerely, l
k.
n M. Crutc
- eld, ief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page t
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SEol SiI o
81040 803E ss4MSEE/[D) h
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Mr. I. R. Fi nf rock, J r.
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G. F. Trowbridge, Esquire Gene Fisher
>l Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 GPU Service Corporation ATTN: Mr. E. G. Wallace Comissioner Licensing Manager New Jersey Department of Energy 260 Cherry Hill Road 101 Comerce 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. Resso, Esquire 248 Washington Street Resident Inspector P. O. Box 1060 c/o V. S. NRC Toms River, New Jersey 08753 P. O. Box 445 Forked River, New Jersey 08731 Joseph W. Ferraro, J r., Esquire Deputy Attorney General Director, Criteria and Standards State of New Jersey Division Department of Law and Public Safety Office of Radiatian Programs l
1100 Raymond Boulevard (ANR-460)
Newark, New Jersey 07012 U. S. Environmental Protection Agency Ocean County Library Washington, D. C.
20460 Brick Township Branch 401 Chambers Bridge Road U. S. Environmental Protection Brick Town, New Jersey 08723 Agency Region II Office Mayor ATTN: EIS C0ORDINATOR Lacey Township 26 Federal Plaza P. O. Box 475 New York, New York 10007 Forked River, New Jersey 08731 l
Comissioner Department of Public Utilities State of New Jersey 101 Comerce Street Newark, New Jersey 07102 1
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EVALUATION OF C0f1 TROL R00 l'150PERAT10fi EVEliTS FOR THE OYSTER CREEK REACTOR (XV-8)
In toiling water reactors t e control rods are moved one at a time (as opposed to gang,,ovement in pressurized water reactors).
During startup and at low i
power (up to a preset value in the 15 to 25 percent of full power range) a rod withdrawal sequence is specified and enforced by procedures, by computer soft. are, or by hard-wired circuitry. Above the preset power level the rod movements _are performed in such a way as to keep the pot.er distribution within the requirements of the limiting conditions for operation and to achieve a particular power shape (e.g., the Haling distribution).
A control rod misoperation event occurs in boiling water reactors when a con-trol rod is moved out of sequence or is moved beyond its usual limits. Two such events - one during startup and low power operation and one at or near full power are usually analyzed.
Rod Misoperation During Startup During startup and low power operation, the rod withdrawal sequence at Oyster Creek is enforced by the Rod Worth Minimizer, a computer-based system that provides motion blocks to the control rod drive system when out of sequence rod motions are attempted.
In the event that the Rod Worth Minimizer is not operable a second operator is rcquired to approve the rod selection and motion befcre the rod is moved. Correctly following the withdrawal sequence consti-
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tutes normal operatica and reactivity additions are designed to permit i
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untroubled increases in power through this range (e.g., no periods so short that a reactor trip occurs before the operator can take action to prevent it).
The probability of a misoperation event occurring during startup and low power operation is small.
fievertheless, a generic analysis of the consequences of such 1
an event has been made and has been reported in fiEDM-23482, " Continuous Control Rod !!ithdrawal Transient in the Startup Range," April 18, 1978.
The analysis
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has been included in the LaSalle County Station Final Safety Analysis Report (Docket 50-341,Section,15.4.1).
The calculation is performed in two steps - first a detailed analysis including multidimensional effects was performed for a rod having a worth in the upper rance of anticipated worths (1.6 percent reactivity change), and then point kinetics calculations were used to extrapolate the results to rod worths expected for out of sequence rods.
Calculations were done with an initial reactor power of one percent of full power because a sensitivity analysis showed that consequences were maximum at this level. Transient termination is assumed
_to ecter by re::i: of the f.DRM scrc:.2 at icw (16 percent) pwcr or the degraded (worstbypasscondition)IRMscram. The withdrawal speed is assumed to be the maximum attainable and rod worths up to 2.5 percent reactivity change were analyzed.
In no case was a peak enthalpy greater than 60 calories per gram encountered. Our acceptance criterion for fuel damage is 170 calories per gram for this event.
On the basis that'the fuel loading and control rod designs for Oyster Creek are essentially the same as those for other later boiling water reactors we i
conclude that this generic analysis applies to the Cyster Creek reactor.
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Rod Misoperation at Pc.er The rod misoperation event at power occurs when the operator cakes a mistake and either withdraws the wrong rod or withdraws a proper rod for too great a l
distance.
In order to taximize the consequences for this ever.t, several con-servative assumptions are made. The cort is assuned to be operating at full power with an assembly or assemblies in the vicinity of the rod to be withdrawn operating at limiting conditions of operation for linear heat generation rate or critical power ratio. The rod to be withdrawn is assumed to be fully in-serted. The existence of a rod pattern which would produce these initial con-ditions implies earlier mistakes by the operator. The core is assumed to be free of xenon and samarium which tends to maximize rod worths. The LPRM de-tectors which would yield the greater response to the event are assumed to be inoperable or bypassed so that protective action is delayed.
It is assumed that the most adverse detector configuration permitted by the Technical Speci-fications is present.-
s.
The calculation is performed with a reactor simulator code and the assumption is made that the neutron flux and thermal flux are in equilibrium. The results of the calculation are used to establish rod block values for_the APV1 channels l
to block rod motion before safety limits on heat generation or critical power l
ratio can be violated.
l The analysis of this event that is currently being used is described in Amendment 76 (Supplement 1) to the Oyster Creek FDSAR submitted on March 25, 1975 (letter from I. R. Finfrock (JCPL) _to A. Giambusso (USNRC)). The procedure to be followed in the future'is described in NED0-24195, " General Electric ne
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I P.eload fuel Application for Oy ter Creek," Section 5.2.1.4 This procedure is
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essentially the same as that v.hich is currently used and is the one that is 4
5 routinely used in the analysis of this event for boiling water reactors.
LI" L'e conclude that the analysis of the at-power rod withdrawal event at Oyster d
Creek is equivalent to that used for current generation boiling water reactors.
This conclusion is based on the following:
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- similar conservative assumptions are made with respect to initial I
conditions, similar calculational tcals are used in the analysis, and
- the acceptance criteria for the event are the same.
I Conclusion Based on the above, we conclude that the analysis of the control rod misoperation event meets current requirements and is therefore acceptable.
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