NLS9100488, Proposed Change 86 to License DPR-46,revising Tech Specs to Remove Rod Sequence Control Sys from Specs & Reduce Low Power Setpoint for Rod Worth Minimizer from 20% to 10% Power

Proposed Change 86 to License DPR-46,revising Tech Specs to Remove Rod Sequence Control Sys from Specs & Reduce Low Power Setpoint for Rod Worth Minimizer from 20% to 10% Power

ML20086C512
Person / Time
Site:	Cooper
Issue date:	11/15/1991
From:	Horn G NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20086C519	List: ML20086C521 NLS9100488, Proposed Change 86 to License DPR-46,revising Tech Specs to Remove Rod Sequence Control Sys from Specs & Reduce Low Power Setpoint for Rod Worth Minimizer from 20% to 10% Power
References
NLS9100488, NUDOCS 9111220282
Download: ML20086C512 (11)

Text

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GENERAL OHICE 9,

• P O. BOX 499. COLUMBUS. NEBRA$KA 686024499 Nebraska Public Power District = ~ - -

TELEPHONE (402) 564 8s61 rA w 0n s u sssi

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NLS9100488 November 15, 1991 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

Proposed Change No. 86 to Technical Specifications Rod Sequence Control System Removal /RWM Setpoint Reduction Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 In accordance with the applicable provisions specified in 10 CFR 50, the Nebraska Public Power District (District) requests that the Cooper Nuclear Station (CNS) Technical Specifications be revised as specified in the attachment. This proposed change removes the Rod Sequence Control System (RSCS) from the CNS Technical Specifications, and reduces the Low Power Setpoint for the Rod Worth Minimizer from 20% to 10% power.

The a:tached contains a description of the proposed change, the attendant 10 CFR 50.92 evaluation, and the CNS Technical Specification pages revised by the institution of this change. This proposed change has been res Lewed by the necessary Safety Review Committees and incorporates all amendments to the CNS Facility Operating License through Amendment 149 issued November 6, 1991.

By copy of this letter and attachment, the appropriate State of Nebraska official is being notified in accordance with 10 CFR 50.91(b)(1). Copies to the NRC Region IV office and the CNS Resident Inspector are also being sent in accordance with 10 CFR 50.4(b)(2). ,

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.U..S. Nucl ear Regulatory Commission Page 2.of 3 November 15, 1991 If you have any._ questions or require any additional information, please contact me.

Sin erely,.

G.h AHorn

. Nuclear Power Group 11anager ,

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' Attachment.

cc: H.R,.Borchert Department of Health -

State of Nebraska NRC Regional Office Region IV Arlington, TX-NRC Resident Inspector

!; Cooper Nuclear Station.

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n O. S. liucicar Regulatory Commission Page 3 of 3 liovember 15, 1991 STATE OF 11EBRASKA)

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PIATTE COUllTY )

G. R. llorn, being first duly sworn, deposes and says that he is an authorized representative of the liebraska Public Power District, a public corporation and political subdivision of the State of fiebraska; that he is duly authorized to s'j r,ubmit this request on behalf of liebraska Public Power District; and that the ~\

s.tatements contained hercin are true to the best of his knowledge and belief.

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Q( II . lic,rn Subscribed in my presence and sworn to '.mfo.. .ne this _I Y- day of

- Ll mM m o A_. , 1991.

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liOTARY PUBLIC EK M M4ff4118sitituts ALOS J. MUOL W Comm. Esa M. 21.1995 h- - -

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REVISED TEr".u1 CAL SPECIFICATIONS j ROD SEQUENCE CONTROL SYSfEM REM dt 9/ ROD WORTH MINIMIZER SETPOINT REDUCTION Revised Pares 18 100 93 101 94 101a 95 216b2 96 216b4  :

97 I. INTRODUCTION The Nebraska Public Power District (District) requests that the NRC approve.the proposed changes to the Cooper Nuclear Station (CNS)

Technical Specifications described below. The proposed changes remove the Rod Sequence Control System (RSCS) from the CNS Technical .,

Specifications.and reduce the Rod Worth Minimizer (RWM) Low Power Setpoint (LPSP)- from its current power level of 20 percent to a power

- level of.10 percent. These changes will enable the District to disable ,

the RSCS at CNS which will-improve reactor startup and controlled i shusdown operations.

• The proposed changes are based on and consistent with the NRC Safety Evaluation Report which documents acceptance of Amendment 17 of General Electric Topical Report NEDE 24011 P A, " General Electric Standard Application for. Reactor Fuel" (GESTAR II).1- This Safety Evaluation '

- generically accepted the removal of the RSCS and reduction of the Rod i Worth Minimizer LPSP from 20 percent to 10 percent. The NRC acceptance is based primarily on~the following:

• The RSCS is redundant to the RWM which the NRC now considers an acceptable and reliable system for control rod pattere control

• Improvements in Control Rod Drap Accident (CRDA) calculation methodology has demonstrated that the CRDA is not of '

significant concern at power levels greater than 10 percent

• An NRC sponsored probability study demonstrates an extremely low prnbability for a CRDA occurring which would exceed the fuel du sge criteria 1

Letter From A. C. Tbndant "RC) to J. S. Charnley (GE) dated December 27, 1987, " Acceptance For Referencing of Licensing Topical Report NEDE 24011-P A, ' General Electric Standard Application For Reactor Fuel,' Revision 8, Amendment 17"

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Attachment to NLS9100488 Page 2 of 8

• Improved control rod movement patterns from 100 percent rod density through 50 percent rod density through use of the Banked Position Withdrawal Sequence (BPWS) which provides reduced maximum control rod worths 8 In accordance with the NRC's safety evaluation accepting Amendment 17 to CESTAR II, the District has reviewed its control rod movement procedures to ensure verification of control rod movement during RWM inoperability, and has incorporated provisions in the proposed Technical Specification changes to minimize reactor startup with the RWM inoperable. A discussion of the procedure review is provided with the description of changes in Section II. Section III provides the significant hazards evaluation.

11. DESCRIPTION OF CHANG 15 A. Technical Specification Page Changes The specific changes proposed to the CNS Technical Specifications are detailed below.

Page-18 - The bases section for Section 2.1.b, "APRM Flux Scram Trip Setting (Refuel or Start 6 ilot Standby Mode)," is revised to delete reference to the RSCS in discussion of systems which back up operating procedures with respect to ensuring uniform control rod pattere.

Page 93 -

Specification a 3.A.2.b is deleted to remove reference to the RSCS. In addition, an editorial change is made '

to correct the spelling of the word " alterations" in Specification 4.3.A.1.

Page 94 - Specification 3,3.A.2.d is revised to remove reference to RSCS. The last paragraph of Specification 3.3.A.2.f is deleted in its entirety, as it is related to operation of the RSCS.

Page 95 - Specifications 3.3.B.3.a and 3.3.B.3.b are deleted in their entirety, as they provide limiting conditions for operation for the RSCS, Correspondingly, -

Specification 4.3.B.3.a. which give the associated surveillance requ;:ements for the RSCS, is also 2

By letter dated February 23, 1988, the NRC approved in Amendment No.

117 to.the CNS Technical Specifications, modifications to the operability requirements for the RSCS and the RWM to allow the Banked Position Withdrawal Sequenco' to be used for rod pattern control between 100 percent and 50 percent rod density.

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4 Attachment to NLS9100488 Page 3 of 8 deleted in its entirety. Specifications 3.3.B.3.c and 4.3.B.3.b are revised to reflect the planned RWM LPSP reduction from 20% power to 10%. Specification 3.3.B.3.c is also revised to add the restriction that only one reactor startup per calendar year is permissible with the RWM inoperabic.

PaBo 96 - Specification 3.3.B.3 e is revised to remove reference to specifications doleted with this proposed change as discussed above, and to reflect the planned RWM LPSP reduction from 20% power to 10% power. Specification 3.3.B.3.f is deleted in its entirety, as this section references use of the individual rod position bypass switches during scram testing which will no longer be needed following disabling and/or removal of the RSCS.

Page 97 - Specification u.3.C.1 is revised to delete reference to Specification 3,3.B.3.a. the RSCS limiting conditions for operation, which is deleted with this proposed change as discussed above. . In addition, the restriction for scram time testing while under 20%

,- power only those control rods "...which were fully withdrawn in the region from 100% rod density to 50%

rod density..." is deleted as this is a restriction related to operation of the RSCS, Page 100 - The first full paragraph on page 100 is deleted in its entirety. This paragraph forms the bases section discussing necessary relaxation of the RSCS restraints _

while performing shutdown margin and control rod drive scram time tests. Additionally, under the bases section 3/4.3.B. "Gnntrol Rod," the reference to the RSCS it deleted.

Page 101 - A number of changes are made to this bases section to remove reference to the RSCS and to reduce the RWM LPSP from 20% power to 10% power. Additionally, in the second-to-last paragraph on the page, the words, "whose qualifications have been reviewed by the NRC" is deleted in reference to individuals other than

-licensed operators whom the District determines are qualiffed as reviewers of rod movements when the RWM 6

is out of service.

Page 101a - Several changes are cmde to this bases section to remove reference to the RSCS and to reduce the RWH LPSP setpoint from 20% power to 10% power.

Additionally, the final two paragraphs under the bases Section 3/4.B.3 are deleted in their entirety. as

Attachment to NLS9100488 Page 4 of 8 i

these paragraphs discuss the surveillances performed l to verify that Group Notch mode restraints are '

enforced during operation of the RSCS.

• Page 216b2- Section 3.22.A.2 and Section 4.22.A.2 are deleted in their entirety, as these sections discuss when and under what conditions the RSCS is allowed to be bypassed while performing certain surveillance testing.

B. Operating Procedures Review In accordanc( vith the~ guidance provided in tne NRC's Safety Evaluation. Report which documents NRC acceptance of Amendment No. IT to CESTAR II, the District has re tiewed the CNS operatin5 procedures to-verify that adequate controls are in place to ensure an independent verification of correct control rod-movement sequence takes p1m e during rod movement with the RWM unavailable.

This review has determined that all CNS procedures which govern  ;

control rod movements require that while the reactor operating at or below the RUM LPSP, a second licensed operator or other '

qualified employee must verify conformance to the correct control rod movement sequence when the RWM is inoperable, III. .SIGNIFICANT HAZARDS DETERMINATION 10 CER 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazards posed by the issuance of the amendment. This evaluation is to be performed with respect to the criteria given in 10 CFR 50.92(c). The following analysis meets these requirements.

Evaluation of this Amendment with Respect t'o 10 C FR 50. 92 The enclosed Technical Specifications change is judged to involve no significant hazards based on the following:

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

Evaluatian

The removal of the RSCS, and the corresponding RWM LPSP reduction from 20% power to 10% power will not involve a significant increase in the probability or consequences of an accident

previously evaluated. The RSCS and the RWM were. originally designed to mitigate the consequences of a CRDA; these systems were not designed to prevent a CRDA from occurring. The probability of a CRDA occurring is a function of the Control Rod e .____..._._,__._a

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NLS9100488 Page 5 of 8 Drive System (CP.DS), which effects movement of the control rods.

Since no hardware changes are being made to the CRDS, the control  ;

rods or the attendant control rod guides, there will be no 1 increase in the probability of a control rod decoupling from its .;

drive or in the probability of a decoupled control rod sticking in  ;

the core. .

i The RSCS was originally designed and installed at earlier vintage Boiling Water Reactors (BWRs). However, it has since been determined that the probability of occurrence of the CRDA is slight, and other, reliable means are employed which effectively minimize the probability of a CRDA occurring in which peak enthalpy values-would exceed the' staff acceptance criteria-of 3 280 cal /gm. In its Safety Evaluation accepting Amendment 17 to  ;

CESTAR II, the NRC referenced a probability study performed by the NRC staff in 1975 to provide a basis for evaluating potential RSCS ,

backfit requirements. This study concluded that for a CRDA to exceed the staff acceptance criteria of 280 cal /gm heat generation in the-peak fuel pellet, the following must ocedr:

"(1) A drive blade disconnect, (2) which is not discovered before rod drop occurs, (3) the blade must '

stick, (4) and not be discovered, (5) the sticking must occur in upper 1/6 of core, (6) the drive must be lowered at least 2 3 feet, (7) an incorrect rod ,

pattern must have been selected and pulled _and, (8) ,

the error not detected, (9) the error muct directly involve the dropped rod and,.(10) the error must i provide an unusually high worth for that rod, (11) the rod blade must unstick and drop, (12) the drop must -

occur at' low power (less than 10%), (13) it must occur when the relevant overall rod pattern is such as to enhance the rod worth (a small fraction of pattern dev61opment time)."

The study conservatively estimated that the probability of a CRDA occurring which exceeds the 280 cal /gm criterion is approximately 1

10 42 per reactor year, a significant margin to an acceptance criteria of 10 4 per reactor year. Since issuance of this study in 1975, approximately ten times the number of reactor years have accumulated without occurrence of a rod drop or even a combination of any two'of the necessary initiating events listed above. Based

- on this data, and-the fact that the probabilityfof a CRDA -r occurring 11s dependent on the CRDS, and not the RSCS or the RWM, the District concludes that the removal of the RSCS and redaction of the RWM LESP does not result in a significant increase of the probability of an accident previously evaluated.

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l Attachment to l NLS9100488 Page 6 of 8 t The RSCS and the RWM were des & 1 ned to mitigate the consequences of i a CRDA. Ilowever, other design and administrative controls are employed which further reduce the possibility of experiencing a '

CRDA which exceeds the 280 cal /gm limit. CNS employs the Banked Position Withdrawal Sequence (BPWS) control rod movement pattern.

The BPWS is a method by which control rods are inserted and withdrawn such that incremental control rod worths are maintained at low values, _thereby mitigating the consequences of the CRDA in the startup and low power operating ranges. 1he BPWS is enforced through the RWM which prevents withdrawing an out-of-sequence control rod more than one notch past the pre-programmed limit.

The CNS procedures which govern control rod movement require that while the reactor is operating at or below the RWM LPSP with the '

RWM inoperable, a second licensed operator or other qualified employce_sha11' independently verify that the proper control rod sequence is being maintained during rod manipulation. '

Additionally, as required by the NRC in the safety evaluation, with this change the District also proposes an administrative limit to minimize reactor atartups with the RWM inoperable. The

- proposed limit is one star tup per calendar year.

-Further, improvements in CRDA analysis methods have indicated that the peak fuel enthalples resulting from a CRDA are significantly lower than previously determined by earlier methods as "

demonstrated by both General Electric and NRC sponsored studies

.(BNL-NUREG'28109, " Thermal llydraulic Effects on Control Drop Accident in a BWR"). These analyses have shown that when above 10% reactor power, a CRDA which exceeds the 280 cal /gm limit cannot occur. Therefore, based on thic and the foregoing discussion, the District has determined that removing the RSCS from operation and reducing the RWM LPSP_from 20% power to 10%

power does not involve a significant increase in the probability -

or consequences of an accident previously evaluated.

Deletion of the portion of the RWM Bases section which indicates L that an other qualified technical plant employee "whose qualification have been reviewed by the NRC" does not' involve a

significant increase in the ,roiqbility or consequences of an accident previously evaluated. It is the District's current position that licensed operators, individuals-qualified as Shift o Technical Advisors, and Reactor Engineering Dcpartment representatives all _ possess the _ qualifications to perform control rod sequence verification activities. Although as stated in 10 CFR 50.36(a), the Bases section does not form part of the L Technical Specifications, it is being revised for accuracy and consistency with the Technical Specifications.

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4 4 Attachment to NLS9100488 Page 7 of 8

2. Does the proposed change create the possibility for a new or different kind of accident from any accident previously evaluated?

Evaluation The RWM and the RSCS were designed only to mitigate the consequences of the CRDA. As discussed above, the District proposes to remove the RSCS and reduce the RWM LPSP. No other  ;

hardware changes or new modes of operation are planned. Likewise, I the proposed change to the RWM bases section does not involve any hardware changes or new mode of operation. Therefore, the District concludes that this change will not create the ,

possibility for a new or different kind of accident from any accident previously evaluated.

3. Does the proposed chpnge create a significant reduction in the mergin of safety?

Evaluation Removal of the RSCS will not create a significant reduction in the margin of safety. The RWM will continue to provide an effective means of supervising control rod movement to ensure that the operators adhere to the correct rod movement sequences. In addition, CNS procedures ensure that during all control red movenants while operating at or below the RWM LP9P with the RWM inoperable, a second licensed operator or other qualified employee verifies that the correct rod sequences are being followed.

Lowering the RWM LPSP from 20% power to 10% power will not reduce

! the margin of safety. Calculations performed by General Electric l and Battelle Pacific Northwest Laboratories have shown that even l with the maximum single control rod position error, and most multiple error patterns, no CRDA can occur which would exceed the acceptance criteria of 280 cal /gm.

l l The NRC has already reviewed and accepted the technical l

justification prepared by General Electric for implementing this I

change. This is documented in the NRC's Safety Evaluation accepting Amendment 17 to GESTAR II. Therefore, the District concludes that removing the RSCS and reducing the RWM LPSP as described above will not create a significant reduction in the l

margin of safety.

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Attachment to '

NLS9100488 Page 8 of 8 IV. CONC 1USION The District has evaluated the proposed changes described above against ,

the criteria given in-10 CFR 50.92(c) in accordance with the requirements of 10 CFR 50.91(a)(1). This evaluation has determined that this proposed change will-D21 1) involve a significant increase in the probability-or consequences of an accident previously evaluated, 2) create the-possibility for a new or different kind of accident from any accident previously evaluated, or 3) create a significant reduction in the margin of safety. Therefore, for the reasons detailed above,'the District requests NRC approval of this Proposed Change No. 86.

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