ML19332D180

Application for Amends to Licenses DPR-42 & DPR-60,deleting cycle-specific Core Operating Limits from Tech Specs & Creating New Core Operating Limits Rept

ML19332D180
Person / Time
Site:	Prairie Island
Issue date:	11/17/1989
From:	Parker T NORTHERN STATES POWER CO.
To:
Shared Package
ML19332D177	List: ML19332D176 ML19332D180 ML19332D181
References
NUDOCS 8911300116
Download: ML19332D180 (6)

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i. UNITED STATES NUCLEAR REGULATORY COMMISSION

, _ NORTHERN STATLS POWER COMPANY i 9

PRAIRIE ISLAND NUCLEAR GENERATING PIANT DOCRET NO. 50-282  ;

50 306 I

.. REQUEST POR AMENDMENT TO OPERATING LICENSES DPR.42 & DPR 60 LICENSE AMENDMENT REQUEST DATED November 17, 1989 j Northern States-Power Company, a Minnesota corporation, requests 4.ithorization .

for changas to Appendix A of the Prairie Island Operating License as shown on the attachments labeled Exhibits A, B, C and D. Exhibit A describes the proposed changes, reasons for the changes, and contains a significant hazards evaluation.

Exhibits B and C are copies of the Prairie Island Technical Specifications incorporating the proposed changes. Exhibit D contsins a sample Core Operating Limits Raport for Unit 1 Cycle 13 and Unit 2 Cycle 13.

This_. letter contains no ristricted or other defense information.  :

NORTHERN ST TF' POWER COMPANY

. By /If&Bf, - I Thomas N Parker Menager Nuclear Support b'ervices

r. , On this/ M ay'of e x d u o /9 k 9 before me a notary public in and for said [

t County, personally a'ppeared Thomas N Parker, Manager Nuclear Support Services, and being first duly sworn acknowledged'that he is authorized to execute this i document on behalf of Northern States Power Company, that he knows the contents thereof and tnat to the best of his knowledge, information, and belief the i statements asde in at are true and that it is not interpreed for delay.  ;

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y Exhibit A l

Prairie Island Nuclear Generating Plant ,

License Amendment. Request Dated November 17, 1989  ;

Evaluation of Proposed Changes to the  !

Technical Specifications Appendix A of Operating License DPR 42 and DPR 60 Pursuant to 10 CPR Part 50, Sections 50.59 and 50.90, the holders of Operating Licanses DPR 42 and DPR 60 hereby propose the following changes to Appendix A, Technical Specifications: ,

Proposed Channes

a. . Core Operating Limit Report A definition of the Core Operating Limits Report has been added to .f Section 1. ~

A new reporti c requirement has been added to Section 6.7 to require the submittal of N,, Core Operating Limits Report prior to the startup of each

. cycle.

b. Heat Flux and Nuclear Enthalpy Rise Hot Channel Factors The Heat Flux and Nuclear Enthalpy Rise hot channel factor limits have

, been moved from Specifications 3.10.B.1 and 3.10 B.2 to the Core Operating Limits Report. Figures TS.3.10 5 and TS.3.10 7 which define the K(Z) and V(2) functions have been moved to the Core Operating Limits Report.

l References to the Core Operating Limits Report have been added to ,

Specifications 3.10.B.1 and 3.10.B.2. ,

c. Axial Flux Difference The axial flux difference limits specified in Figure TS.3.10 6 and the axial flux difference target band requirenents of Specifications 3.10 B.4,  :

3.10 B.6, 3.10.B.8 and 3.10.B 9 have been moved to the Core Operating Limits Report.

References to the Core Operating Limits Report have been added to Specifications 3.10.B.4 and 3.10.B.6.

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d. Shutdown and Control Bank Insertion Limits The shutdown bank withdrawal requirements of Specification 3.10.D.1 have ,

been moved to the Core Operating Limits Report. . The control bank I insertion requirements of Specifications 3.10.D.2, 3.10.D.3, 3.10.G.3 and 3.10.G.4 and Figures TS.3.10 2, TS.3.10 3 and TS.3.10 4 have been moved to the Core Operating Limits Report.

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References to the Core Operating Limits Report have been added to Specifications 3.10.D.1, 3.10.D.2, 3.1.F.3.a. 3.10.G.3 and 3.10.G.4.

e. Reactor Coolant System Flow Limit The reactor coolant system flow limit in Specification 3.10.J has been E

moved to the Core Operating Limits Report and has been replaced with a ,

reference to the Core Operating Limits Report.

, f. Table of Contents  :

The Table of Contents and the List of Figures have been modified, as shown '

i in Exhibit B, to reflect the changes described above,

g. Bases Modifications The bases have been modified, as shown in Exhibit B, to remove cycle specific variables and replace them with references to the Core Operating l Limits Report.

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Reason For Chances b Generic Letter 88-16, dated October 4, 1988, was issued to encourage licensees to prepare' changes to Technical Specifications related to cycle specific j parameters. The generic letter provided guidance for the relocation of 6 certain cycle dependent core operating limits from the Technical Specifications. This would allow changes to the values of these cycle-dependent core operating limits without prior NRC approval (i.e., license s amendment), so long as an NRC approved methodology for the calculations is '

followed, r

L The proposed Technical Specification changes, which relocate cycle specific  !

core operating limits from the Technical Specifications to the Core Operating -

Limits Report, are being submitted in accordance with the guidance provided in l_ Generic Letter 88 16. The proposed changes reference the Core Operating l

Limits Report for cycle specific core operating limits and ensure that the cycle specific core operating limits are maintained within the limits of the -

Core Operating Limits Report. The proposed changes to the administrative controls section ensure that the calculation of the core operating limits ,

proposed for inclusion in the Core Operating Limits Report will be performed in accordance with NRC approved methodologies.

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.. . zum e res : er s Reactor coolant system flow is measured each cycle. Significant margin exists between the actual reactor coolant system flow aad the reactor coolant system ,

flow limit being relocated to the Core Operating Limits Report. Relocation of the reactor coolant system flow limit to the Core Operating 1.imits Report will provide us with the ability to utilize some of the margin between the actual reactor coolant system flow and the current flow limit to gain additional margin in other core operating limits.

l The proposed changes are consistent with the requirements of 10 CTR 50.36 and the NRC Staff's proposed policy for improving Technical Specifications, delineated in SECY.86 10. " Recommendations for Improving TS". The proposed policy allows process variables, such as core operational limits, to be controlled by specifying them numerically in the Technical Specifications or

by specifying the method of calculating their numerical values if the NRC Staff finds that the correct limits will be followed in operating the plant.

The proposed revision references the NRC approved calculation methodologies.

The development of cycle specific core operating limitc will continue to be performed by the referenced methodologies which have been accepted by the NRC.

The proposed changes to the Technical Specifications are also considered to be improvements and arn consistent with the NRC stated policy for improving Technical Specifications (52 FR 3788 February 6,1987).

Safety Evaluation and Determination of Significant Hazards Considerations The current Technical Specification method of controlling reactor physics parameters to assure conformance to 10 CPR 50.36 is to specify the values determined to be within the acceptance criteria using an NRC-approved calculation methodology. As previously discussed, tbc methodologies for calculating these parameters have been reviewed and approved by the NRC and are consistent with the applicable limits in the Updated Safety Analysis Report.

The removal of cycle & pendent core operating limits from the Technical Specifications has no impect upon plant operation or safety. No safety-related equipment, safety function, or plant operations will be altered as a result of the proposed changes. Since the applicable Updated Safety Analysis Report limits will be maintained and the Technical Specifications will continue to require operation within the core operational limits calculated by NRC approved methodologies, this proposed change is administrative in nature.

Appropriate actions to be taken if limits are violated will also remain in the Technical Specifications.

The proposed changes will contro' the cycle-specific parameters within the I acceptance criteria and assure conformance to 10 CPR 50.36 by using the approved methodology instead of specifying Technical Specification values.

The Core Operating Limits Report will document the specific core operating limits resulting from calculations performed utilizing the NRC approved methodologies specified in the Technical Specifications. Therefore, the proposed change is in conformance with the requirements of 10 CFR 50.36, 1

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From cycle to cycle, the. Core Operating Limits Report will be revised such that the appropriate core operating limits for the applicable cycle will apply. Technical Specifications will not be changed.

The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10

' CPR Part 50, Section 50.91 using the standards provided in Section 50.92.

This analysis is provided below:

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

The removal of cycle specific core operating limits from the Prairie Island Technical Specifications has no influence or impact on the probability or consequences of any accident previously evaluated. The cycleispecific core operating limits, although not in Technical Specifications, will be followed during the operation of the plant. The propcted_ amendment still requires the same actions to be taken when or if limits are exceeded as is required by current Technical Specifications.

Each accident analysis addressed in the 11pdated Safety Analysis Report will be examined with respect to changes in cycle. dependent parameters, which are obtained from application of the NRC-approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses. This examination, which will be performed per Jequirements of 10 CFR 50.59, ensures that future reloads will not involve a=significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident vreviousiv analyzed.

As stated above, the removal of the cycle specific core operating limits from the Prairie Island Technical Specifications has no influence or impact, nor does it contribute in any way to the probability or consequences of an accident. No safety related equipment, safety function, or plant operations will be altered as a result of the proposed changes. The cycle-specific core operating limits will be calculated using the NRC. approved methods and submitted to the NRC to allow the Staff to continue to trend the values of these limits. The Technical-Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

Therefore, the proposed amendment does not in any way create the possibility of a new or different kind of accident from any accident

>-- previously evaluated.

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'4 hhibit A Pas) $ st S e i I

3. The proposed amendment will not involve a significant reduction in the [

martin of saferv. -

n The margin'of safety is not affected by the removal of cycle specific core L operating limits from the Technical Specifications. The margin of safety ,

provided by the current Technical Specifications remains unchanged.  ;

j Appropriate measures exist to control the values of the cycle-specific limits. The proposed license amendment continues to require operation i IJ within the core limits obtained from NRC approved reload design  !

i methodologies. The actions to be taken when or if limits are violated I remain unchanged.

Therefore, the proposed changes are administrative in nature and do not i impact the operation of the plant in a manner that involves a reduction in I the margin of safety.  !

L The Commission has provided guidance concerning the application of the ,

j, standards in 10 CFR 50.92 for determining whether a significant hazards r ,

consideration exists by providing certain examples of amendments that will  ;

.- likely be found to involve no significant hazards considerations. These l' examples were published in the Federal Register on March 6, 1986.

The changes to the Frairie Island Techni:a1 Specifications proposed in this o amendment request are equivalent to NRC example (1), because they involve purely administrative changes intended only to clarify existing Technical Specifications. Based on this guidance and the reasons discussed above, we have concluded that the proposed change does not involve a significant hazards ,

consideration.

Environmental Asseasment This license amendment request does not change effluent types or total

. effluent amounts nor does it involve an increase in power level. Therefore, this change will not result in any significant environmental impact. l l

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