ML20197H840

Application for Amends to Licenses DPR-42 & DPR-60,revising Tech Spec 3.10.G to Allow Continued Operation for 72 H for Diagnosis & Repair

ML20197H840
Person / Time
Site:	Prairie Island
Issue date:	11/14/1990
From:	Parker T NORTHERN STATES POWER CO.
To:
Shared Package
ML20197H838	List: ML20197H832 ML20197H840 ML20197H850
References
NUDOCS 9011200203
Download: ML20197H840 (6)

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UNITED STATES NUCLEAR RECULATORY COMMISSION 1

NORTHERN STATES POWER COMPANY PRAIRIE ISLAND. NUCLEAR GENERATING PLANT DOCKET No. 50 282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60

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LICENSE AMENDMENT REQUEST DATED November 14, 1990 Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown on the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and a significant hazards evaluation. Exhibits B and C are copies of_the Prairie Island Technical Specifications incorporating the proposed changes.

This letter contains no restricted or other defense information.

NORTHERN STA ES POWE OMPANY By Thomas M Parker Manager' '

Nuclear Support Services On this ay of b ed ," M .>efore me a notary public in and for said County, personally a)peared Thomas M Parker, Manager Nuclear Support Services,-

and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the I statements made in it are t u and that it is not interposed for delay.

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Prairie Island Nuclear-Generating Plant  !

License Amendment Request Dated November 14, 1990 ,

Evaluation of Proposed Changes to the [

Technical Specifications Appendix A of_ j' Operating License DPR-42 and DPR-60 E j Pursuant to 10 CFR.Part 50, Sections 50.59 and 50.90, the holders of" Operating g Licenses DPR-42'and DPR 60 hereby propose the following changes to Appendix A; <

L Technical Specifications: .

1. Rod' Control Operability Changes Proposed Changes L

. .c Revise specification 3.10.G 1-as shown-on page TS.3.10-7 in Exhibit B to D 4 ' clarify that this specification is applicable only to a control rod which 4 cannot be, moved as a result of excessive . friction or mechanical  ;

, interference and to remove the.eight hour time allowance. :i Revise specifications 3.10.G.3 and 3.10.G.4 as shown on page TS.3.10 7'in l

. Exhibit B to. clarify that these specifications are applicable ~to both:  :!

e inoperable and immovable control rods-

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Addjnowspecification3.10.G6,asshownonpageTS.3.10-7inExhibitB, which specifies the-actions-to be.taken.in. response to control rods which  ;

e :are immovable due to'an electrical problem in the rod control'~ system. .g Revise-the heading to Section 3.10.G on pages_TS.3.'10-7 and~B.3.10-9, and Table.of Contents pages TS-iv and TS x, as shown in Exhibit'B, to clarify the. applicability of Section 3.10.0. ,

Incorporat'e bases for the proposed changes into the Control Rod and Power

, Distribution Limits specification bases as shown on pages~B.3.10 9 and j' iB.3.10-10 in Exhibit B.

Reason For Chances. O Current. specification 3.10.G.1 specifies that a control rod which cannot~

be moved by its drive mechanism and cannot 'be corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> be -

declared inoperable. Specification 3.10.G.2 specifies that-the reactor shall be brought to the hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> should more than one inoperable rod be discovered during power operations. These  !

., current specifications fail to recognize the greater significance of rods being-immovable due to mechanical interference. Acknowledgement of the lesser significance of electrically immovable rods would allow additional i

. time for diagnosis and repair.of malfunctioning equipment while maintaining safe operation of the plant.

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Exhibit A 1 Page 2 of 5 The proposed changes provide distinct action statements.for immovable rods that are more; consistent with the significance of the malfunctions. A rod that is immovable due to excessive friction or mechanical interference is more significant than a rod that cannot be stepped due to an electrical  ;

malfunction, but remains.trippable. Distinguishing between these types _of i 7 malfunctions will allow an appropriate-time period to complete corrective action commensurate with the significance of the malfunction. The proposed changes to the Technical Specifications would allow continued operation-for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for diagnosis and repair, with one or more control U

rods immovable due to an electrical problem in the rod control system.

Extending the allowed out of. service time for rods immovable due to rod control system' failures-would accomplish the following:

1. Allow sufficient time to evaluate the nature of the failure, to develop _ -l a systematic work plan and to perform troubleshooting activities in a: 1

. deliberated and systematic manner. This would reduce the chances of-I dropping a rod or tripping the reactor during the troubleshooting _

activity, i

2. Some of the-failures will' randomly occur during nights and weekends. {

The proposed changes would allow time for the most experienced people 1 available to travel to the plant, to plan and perform the work. The 1 extended . allowed out' of service -time would also provide additional time j "for. consultation-with vendor personnel as necessary. ]

h 13.IIn.th'e event of parts u'ava11 n ability, the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> would provide A additional time to procure the required parts. 'j

14. ' Plant : shut'down without the use of some or all of the control rods is an i

. abnormal operation which could lead to unforseen complications and. j which should-be avoided if possible. The proposed = changes would reduce '

.the potential for plant shutdowns resulting from rod control system failures. g Amendment 27 to the Wolf Creek ' Unit 1 Technical Specifications, dated March 6, 1989, incorporated revised actions to be taken in response to control rods which are electrically immovable but still tri i pable. The-  ;

actions proposed for the Prairie Island Technical Specificat ions for 1 electrically immovable rods are similar to those approved fc r Wolf Creek.  ;

The Wolf Creek specification states that with one or-more control rods electrically inoperable, but trippabic, restore the inoperable rods =to l operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot standby within the next 6  !

hours. The differences between the approved Wolf Creek specification and the actions proposed by this amendment request are discussed below: )

1. The proposed Prairie Island specification requires the rod insertion-limits of specification 3.10.D be verified within one hour of the discovery of electrically immovable, but trippable, rods. Verification /

that control rod position is within the rod insertion limits will [

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Exhibit A Page 3 of 5 g i

provide' assurance of' adequate shutdown margin. There are no 'j requirements in the_ Wolf Creek specification for the verification of '

-shutdown margin or rod insertion limits.

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-2.'As stated above, the' Wolf Creek specification requires the unit to be taken to hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if the rods'are not operable at the j end of the 72~ hour allowed out of service time. At the end of the 72 <

hour-allowed out of service time, the proposed Prairie Island

. specifications declare any remaining electrically immovable rods',

inoperable, and applies the requirements of specification 3.10.G.2.

Specification 3.10.G.2' requires the unit to be taken to hot shutdown

• within 6. hours if more than one control rod is inoperable. ,

The current Prairie Island Technical Specifications allow continued'  !

. operation with one inoperable rod, as defined by specification 3.10 G.I. . Continued operation with one inoperable rod is only allowed 1 if the additional. restrictions of specifications 3.10 G.3, 3.10-G.4 and 3.10 G.5 are met.

- Specifications-3.10.G.3 and 3.10.G.4 invoke more  !

restrictive rod. insertion limits based on the position of the- j 11noperable, rods. These more restrictive rod insertion limits could ,

result in reduced power: operation under certain conditions.

Specification3.10.G.5 requires that if power operation is continued with an inoperable rod,1 the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by j analysis within 30 days. If the analysis results in a more limiting. .

hypothetical transient than the' cases reported in.the safety analysis,-

reactor power is reduced to a level consistent'with the safety analysis, y k The proposed action statements would allow continued power ' operation '

with-one electrically immovable, but trippable, control rod as long as ~

the' requirements of specifications 3.10.G.3, 3.10.G.4 and 3.10.G.5 are met. This is c'onsistent with the current Prairie Island Technical-  !

LSpecification requirements, which would allow continued onoration with one electrically inoperable control rod. We~believe continued power operation with an electrically immovable, but trippable, rod is justified by the additional restrics 'nns invoked .by specifications 3.10 G.3, 3.10.G.4 and 3.10.G.S. Furtner justification .is provided by the:NRG approved Prairie Island reload analysis which is based on the current Technical Specifications and which is performed assuming two stuck rods. The rod insertion limits invoked by' specifications-3.10.G.3 and 3.'10.G.4 bound the reload analysis.

• The Prairie Island llot Shutdown condition is equivalent to the llot

. Standby condition utilized by the Wolf Creek specification and by the

' Standard Technical Specifications.

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Exhibit A Page 4 of 5 i

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' Safety Evaluation and Determination of Sienificant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine.whether they constitute a significant hazards-consideration as

', . required by 10 CFR Part 50, Section 50.91 using the standards provided in h Section 50.92. This analysis is provided below:

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1. The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previousiv evaluated.

The proposed changes do not affect the ability of the Control Rod Drive System to perform its intended safety function (reactor trip). The design of the Control Red Drive System hssures isolation of the elements. required to insure reactor trip from the' Rod Control System.

Extending the allowed-out of service time associatad-with electronic / electrical malfunctions of the Rod Control System is

- acceptable, since the safety function of the Control Rod Drive System (reactor' trip) is notfaffected by the change. Because the reactor trip function is not affected, the conclusions in the Prairie Island Updated Safety Analysis Report remain valid.

-Therefore, based on the conclusions of the above discussion, the proposed changes will not. involve a significant increase in the fprobabilityaor consequences of an accident previously' evaluated.

L2. The proposed. amendment.will not create the possibility.of a new or different kind of accident from any accident oreviousiv analyzed.  ;

1 There are no new failure modes or mechanisms associated with the t proposed changes. The proposed changes do not involve any modification  !

in theLoperational limits or physical design of the' involved systems.

The change merely allows an extended time period for-the diagnosis and repair of Rod-control-System failures,.thus' reducing the probability of.

a plant transient because of' insufficient time for proper corrective

. action-or a hurried diagnosis.

As discussed above, the proposed changes do not result in any  ;

significant change in the configuration'of the plant, equipment design  !

.or equipment use nor do they require any change in the accident analysis methodology. .Therefore, no different type of accident is created. No safety analyses are affected. The accident analyses presented in the-Updated Safety Analysis Report remain bounding.

3,'The proposed amendment will not involve a significant reduction in the marcin of safety.

The proposed < changes do not affect any Technical Specification margin of' safety. The proposed changes allow appropriate actions commensurate with the significance of Control Rod Drive or Rod Control System L malfunctions, while not requiring plant transients in response to l

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Exhibit A Page 5 of 5 I malfunctions that do not affect the capability of the control Rod Drive

~ System to perform its safety function.- Therefore, the proposed changes-will not result in a significant reduction in the plant's-margin of #

safety. 4 i

7 The' Commission has provided guidance concerning the application of the ,

standards in'10 CFR 50.92 for-determining whether a significant hazards consideration exists by providing certain examples of amendments that will. t likely be found to involve no significant hazards considerations. These examplas were published in theLFederal' Register on March 6, 1986. , ~!

The changes.to the Prairie. Island Technical Specifications proposed above o .are equivalent to NRC example (vi)', because they involve changes which

either may result in some increase to the probability or consequences of a -

previously-analyzed accident'or may reduce in some way a safety margin, i

'but where the results-of the change are clearly within all.ac'ceptable criteria with-respect to the system.or component specified in the Standard Review Plan. Based on this' guidance and.the reasons discussed above, we' thave concluded-that the proposed changes do not-involve a significant hazards consideration. ,

Environmental Assessment. l

.This licensefamendment-request does rat change effluent types or. total  ;

5 effluent amounts nor does it involve ~an increase in power level. Therefore, c.

this change;will not result'in any significant environmental impact.

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