ML20237E884

Application for Amend to License DPR-46 Consisting of Proposed Change 54,revising Tech Specs to Modify Operating Requirements for Rod Sequence Control Sys & Rod Worth Minimizer.Fee Paid

ML20237E884
Person / Time
Site:	Cooper
Issue date:	12/22/1987
From:	Kuncl L NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20237E886	List: ML20237E884 ML20237E899
References
NUDOCS 8712290242
Download: ML20237E884 (7)

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December 22, 1987 U.S. Nuclear Regulatory Commission Document ~ Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

Proposed Change No. 54 to - Technical Specifications Cooper Nuclear Station' N R C Docket No. 50-298, D PR-46 In accordance with the ap plicable provisions specified in 10CFR50, Nebraska Public Power District requests that Technical Specifications for Cooper Nuclear Station (C NS) be' revised to modify the operating requirements for the Rod Sequence Control System (RSCS) and Rod Worth Minimizer (RWM).- The changes ,

will allow the Banked Position Withdrawal' Sequence to be followed I for the first 50% of control rod withdrawals resulting in lower maximum rod worths during startu p. Several administrative changes in the Technical Specification Bases are also requested.

A discussion and the applicable revised Technical Specification pages are contained in Attachment 1. The modifications to the Technical Specifications within this proposed change have been evaluated with respect to the requirements of 10C FR50.92. The results of the evaluations are also included in the attachment.

This proposed change incorporates all amendments to the CNS Facility O perating License throug h Amendment 112 issued October 27, 1987. By copy of this letter and attachment the appropriate State of Nebraska official is being notified in accordance with 10C FR50.91(b).

This change has been reviewed by the necessary Safety Review Committees and payment of $150 is submitted in accordance with 10 C F R170.12.

8712290242 871222 hoo\

PDR ADOCK 05000298 p.00 r 'g getJw uco' g.,.

Page 2' .I December 22, 1987 l

In addition to the: signed. original, 37 copies 'are also submitted for ~your use. Copies to the NRC Region IV Office and theLCNS.

Resident Inspector are also being sent in accordance : with 10 C F R 50.4(b)(2) . - Should you have any questions .or require additional .information, please contact me.

Sincere y, 4

L . G . K u nci Vice-President - N uclear L G K/grs:dmr3/4 -

Attach ment cc: H. R. Borchert ..

Department of Health State of Nebraska N R C . Regional Office Region IV Arlington, TX l N RC Resident Inspector Office Cooper Nuclear Station i

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.! Page 3 December 22, 1987 ,

ST ATE OF NEBRASKA)

)ss PLATTE COUNTY )

L. G. Kuncl, being first duly sworn, deposes and says that he is an authorized repre::entative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to sLbmit this request on behalf of Nebraska Public Power District; and that the statements contained herein are true to the best of his knowledge and belief.

L . 6) K u ncl S u bscribe in m resence and sworn to before me this day of 01 % O , 1987.

- N O T A R Y P U B,LIC

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conEEN M. KUTA Ny Ceest Em Aug 4,195 i

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• e < . - Attachmsnt II Revised Technical' Specifications for RSCS and RWM Operability.-

Revised Pages: 95 . .

100 95a (deleted) 101 96 101a:

99 104

Reference:

1) NRC Letter from Cecil O. Thomas to J. S. Charnley. " Acceptance

= for Referencing of Licensing. Topical Report NEDE-240ll-P-A.

' General Electric Standard Application for Reactor Fuel ,

Revision 6, Amendment 12," October. 11,.1985.

2) General Electric Service Information Letter ' No. - 316 Reduced Notch Worth Procedure, November 1979.

In order'to minimize control. rod reactivity worths' associated with the Control' Rod Drop Accident (CRDA), BWR control- rods are; pulled in specific; patterns during startup that are supervised.by. control ~ systems: such as the Rod Worth Minimizer (RWM) computer program or the hard wired Rod Sequence Control System-(RSCS). These rod patterns fall.into two classes of operation'for BWR. plants and restrict rod motions:to ensure that a postulated'CRDA will'not result'in peak fuel enthalpies in excess. of 280 calories per gram.for; the Lentire range of plant operations at.d core ' exposures.

Some BWR/4s, including Cooper Nuclear Station, are Group ^ Notch Rod Sequence Control System (CNRSCS) plants which utilize the'RSCS.with the RUM as.a backup to prevent high control rod reactivity worths. During 'startup ' prior to reaching the 50 percent rod density pattern, each of the control rods can be withdrawn in a specified sequence from its full in to its full out. position in one continuous. motion. Once the 50 percent' rod density pattern (checkerboard) is reached, further rod withdrawals take place in defined banks one notch at a time until a preset rated reactor power level is reached at which time RSCS constraints on rod patterns for CRDA consequences are not required. Above this minimum power setpoint the consequences of Ja' CRDA are mitigated by the presence of steam voids in the coolant which provided an additional negative reactivity feedback mechanism and the CRDA is of minimal safety fconcern.

Other BWR plants utilize a Bank'ed Position Withdraw Sequence "(BPWS) which restricts the first 50 percent of the rods withdrawn so ' that they 'are scepped out in a specified (banked) pattern rather than in one-continuous motion. This method of withdrawal is used for rod movement beyond the 50 percent density as well. The RSCS is not used in the sequence and rod motion is controlled by the RWM alone. Due to the use of the banked positions for. the first 50% of the rod withdrawal (pre-checkerboard), the BPWS 1 results in lower peak' rod reactivity worths for the CRDA than the GNRSCS pattern. Beyond 50% of rod withdrawal (post-checkerboard) the GNRSCS patterns provide a somewhat lower worth than the corresponding BPWS patterns.

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'.I For GNRSCS plants, the higher rod worths for the pre-checkerboard pattern result in an expensive and time-consuming analysis for each core reload of the Reference Loading Pattern to predict the results of a potential CRDA. At -

times the core has to be redesigned solely to meet. the CRDA limits with the resulting reload core configuration less than optimal for fuel cycle cost considerations. For the BPWS plants the rod worths are sufficiently low both in the first 50% rod withdrawal and beyond that statistical studies conclude with a high degree of confidence-that the CRDA will not exceed the required limits and that reload analyses for the accident are not required.for core reinads. If the GNRSCS plants could use the BPWS for the first.50% and the group notch mode for post-checkerboard rod movement (since the CNRSCS rod worths are lower), then they could take credit for the BPWS statistical analysis and eliminate the need for a reload CRDA analysis. Reference 1 evaluated the use of and credit for the BPWS in the pre checkerboard region for those plants which utilize the GNRSCS and found it to be acceptable.

Reference 1 went on to state "that it is preferable for the CNRSCS plants to have the improved pattern control of the BPWS as monitored by the RWM for.the first 50 percent of withdrawal. The ~ GNRSCS group notch ' mode should be retained beyond 50 percent. Plants making the change will be able to take I

credit for the statistical analysis of the CRDA and will not have to analyze the event for reloads." Reference 1 further stated that plants wishing. to make this change are to change Technical Specifications as required and to j

indicate that BPWS patterns will be enforced during applicable operations.

Cooper Nuclear Station currently u'tilizes the Reduced Notch Worth Procedure (Reference 2), which is basically an extension of BPWS, for the RWM input. As such, the RWM imposes more severe constraints on rod motion during the first 50% than does the RSCS. Plant startup and shutdown rod motion' operations will not be significantly changed by adapting the BPWS for the . first 50% rod density since the Reduced Notch Worth Procedure is being followed as recommended in Reference 2.

Consequently, the District requests changes be made to the CNS Technical Specifications to require adherence to a BPWS for RWM input and to delete RSCS operability requirements for the withdrawal of the first 50 percent of the rods from the core. From 50 percent rod density to 20 percent rated thermal power rods will continue to be withdrawn in the group notch mode as controlled by the RSCS. With the more restrictive (lowe.r maximum rod worth) BPWS pattern in place being supervised by the RWM computer system, the RSCS can be inoperable while the BPWS is in effect. Specifically, the following changes are proposed: ~

1. On Page 95 amend Sections 3.3.B.3a and 3.3.B.3b to require the RSCS to be l operable from the 50% rod density to 20 percent of rated thermal power region.

2. On Page 95 revise the surveillance requirements for the RSCS by deleting the sequence portion of the testing and reformatting the group notch j portion of the checks for startup and shutdown.

3. Repaginate Specification 4.3.B.3.b to Page 95 and delete Page 95a in its entirety. I l

4. On Page 96 amend Specification 4.3.3.3.b.11 to verify the correctness of the BPWS input to the RkH computer.

5. Amend the Bases on Page 101a to discuss the use of the BPWS.

6. Add Reference No. 4 to Page 104 on Reduced Notch Worth Procedure.

7. Finally, the District requests to amend the bases on Pages 99, 100, and 101 to change references of the Final Safety Analysis Report to the l Updated Safety Analysis Report. Also on Page 101, a no longer applicable l

USAR reference has been deleted.

I Evaluation of this Revision with Respect to 10CFR50.92 g A. The encic ed Technical Specification change is j udged to involve no significant hazards based on the following:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluetion: I

a. The proposed amendment would allow the. Rod Worth Minimizer (RkM), enforcing the Banked Position Withdrawal Sequence, to take -the place of the Rod Sequence Control System (RSCS) in controlling the allowed control rod patterus and rod reactivity worths during the period from all rods inserted into the core to 50% of the rods withdrawn. The use of the RkH enforcing the, Banked Position Withdrawal Sequence (BPWS) has been evaluated to lessen the consequences of a Control Rod Drop Accident over the present use of the RSCS during the first 50% of rod withdrawal because the BPWS bank positions result in lower peak rod reactivity worths. Thus, the use of the RkH utilizing the BPWS during the first 50% of rod withdrawal will reduce the consequences of any Control Rod Drop Accident (CRDA). The probability of the CRDA is not affected by the change ao current control rod to drive mechanism coupling requirements and surveillance remain .in effect. The proposed amendment )

affects no other evaluated accident scenarios. ,

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b. The proposed amendment also contain various changes to the {

bases section that are administrative in nature, namely changing reference from the Final Safety Analysis Report (FSAR) {'

to the Updated Safety Analysis Report (USAR) which is periodically revised to reflect current plant configuration, analysis and operation. Additionally, an out-of-date reference is being deleted. These administrative changes do not affect any of the assumptions or analysis contained in the Safety j Analysis Report and hence do not increase the probability or consequences of any previous evaluated accident.

2. Does the proposed licensa amendment create the possibility for a new or difference kind of accident from any accident previously evaluated?

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Evaluation: ,

a. The proposed amendment places a new restriction on rod' motion - i

'during the first 50% of red withdrawal to re' duce ' peak' rod . .

reactivity and miti;, ate the effects of a hypothetical- Control )

Rod Drop Accident.' No new mode-of plant operation is created

~'or made possible by the change ~which ^only affects. control rod j

, movement blocks. The . proposed -. amendment (does not create the ,

. possibility _ for a. new or different kind 'of. accident ' froml any -

i previously evaluated.H i

b. The proposed . administrative changes ' have no , effect on plant-operation or its underlining analysis,: so will not create the-possibility for a new or different kind of accident.

3. Does . the proposed amendment involve a significant : reduction in' a margin 'of safety? -

a. As explained above,' the proposed amendment places additionali I

rod motion.r'estraints in the'first 50% of rod ' withdrawal' that '

results.in lower peak: rod reactivity worths. This will' result' in a lower peak enthalpy reached in the fue1~in the' event'of a-Control Rod Drop Accident and will add' additional margin to the:

280 calories / gram limit. The proposed change does.not involve a significant reduction in~a margin of safety.

b. The proposed administrative changes have :no effect on plant operations or the value of its safety limits and involves no .

significant reduction in a margin of safety.

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