ML20011F433

Application for Amend to License NPF-49,revising Tech Specs Re Relocation of cycle-specific Core Operating Limits for Unit from Tech Specs to Core Operating Limits Rept

ML20011F433
Person / Time
Site:	Millstone
Issue date:	02/21/1990
From:	Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20011F434	List: ML20011F433 ML20011F435 ML20011F436
References
B13435, GL-88-16, NUDOCS 9003060029
Download: ML20011F433 (8)

Text

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i February 21, 1990 i Docket No. 50-423 B13435 i L .

L U.S. Nuclear Regulatory Commission  ;

l . Attention: Document Control Desk l Washington, DC 20555 i Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 l Proposed Revision to Technical Specifications j Removal of Cycle Specific Parameter j Limits from Technical Soecifications  !

Pursuant to 10CFR50.90,. Northeast Nuclear Energy Company (NNECO) hereby 4 proposes to amend operating license NPF-49 by incorporating the changes ]

identified in Attachment 2 into the Technical Specifications of Millstone Unit  ;

No. 3.

P Backoround l- Generic Letter 88-16, dated October 4, 1988, was issued to encourage licensees i to prepare changes to Technical Specifications related to cycle-specific l parameters. These Technical Specification changes will relocate cycle-specific parameter limits from Technical Specifications to the Core Operating Limits Report (COLR), Presently the parameter limits in the Millstone Unit )

l. No. 3 Technical Specifications are calculated using NRC-approved methodolo- 1 L gies. These limits are evaluated for every reload cycle and may be revised j periodically as appropriate to reflect changes to cycle-specific variables. l This is an administrative burden on both the NRC and NNECO. j The generic letter provided guidance to allow relocation of certain cycle 1

-dependent core operating limits from the Millstone Unit No. 3 Technical Specifications. This would allow changes to the values of core operating limits without prior approval. (i.e., license amendment) by the NRC, provided l an _ NRC-approved methodology for the parameter limit calculation is followed. H

(; Thus, future core reloads and other revisions will require a safety review in j

) accordance with the requirements of 10CFR50.59 instead of a prior submittal to the NRC.

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Currently, each parameter limit proposed in the COLR utilizes the approved methodologies identified in the revised Administrative Controls section of this license amendment request. Millstone Unit No. 3 will use these methodol-ogies when performing core reload designs and when any other revisions are made, j 9003060029 900221 -

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U.S. Nuclear Regulatory Commission B13435/Page 2 February 21, 1990 Description of the Proposed Chanaes The proposed Technical Specification changes concern the relocation of several cycle specific core operating limits for Millstone Unit No. 3 from the Technical Specifications to the COLR. A new definition of the COLR is being added to the Technical Specificationc. Aditionally, certain individual Technical Specifications will be moa1fied to note the cycle-specific parameter limits arr. contained in the COLR. A COLR paragraph is being added to the Administrative Controls Section which will replace the Radial Peaking Factor Limit Report.

The cycle-specific parameter limits proposed for relocation to the COLR as '

part of this license amendment request include:

3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limits (Four Loop and Three Loop) 3/4.2.1.1 Axial Flux Difference - Four Loop 3/4.2.1.2 Axial Flux Difference - Three Loop 3/4.2.2.1 Heat Flux Hot Channel Factor - Four Loop 3/4.2.2.2 Heat Flux Hot Channel Factor - Three Loop 3/4.2.3.1 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor

- Four loop 3/4.2.3.2 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor

- Three Loop Attachment 1 provides a description of the proposed changes and Attachnent 3 is a sample COLR based on data for the current fuel cycle.

Safety Assessment The current Technical Specification method of controlling reactor physics parameters to assure performance to 10 CFR 50.36 (which requires the lowest functional performance levels acceptable for continued safe operation) is to specify the values determined to be within the acceptance criteria using an NRC-approved calculation methodology. As previously discussed, the methodolo-gies for calculating these parameter limits have been reviewed and approved b.y the NRC and are consistent with the applicable limits in the Final Safety g

Analysis Report (FSAR).

The removal of cycle-dependent variables from the Technical Specifications has no impact upon plant operation or safety. No safety-related equipment, safety function, or plant operations will be altered as a result of these proposed changes. Since the applicable FSAR limits will be maintained and the Technical Specifications will continue to require operation within the core operational limits calculated by these NRC-approved methodologies, the proposed changes are administrative in nature. Appropriate actions to be taken if limits are violated will also remain in the Technical Specifications.

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U.S. Nuclear Regulatory Commission l B13435/Page 3 February 21, 1990  ;

These proposed changes will control the cycle-specific parameters within the acceptance criteria and assure conformance to 10 CFR 50.36 by using the approved methodology instead of specifying Technical Specification values.

The COLR will document the specific parameter limits resulting from Millstone Unit No. 3 calculations, including mid cycle or other revisions to parameter .

-lues. Therefore, the proposed change is in conformance with the require-ments of 10 CFR 50.36. ,

Any changes to the COLR wili be made in accordance with the provisions of 10 CFR 50.59. From cycle to cycle, the COLR will be revised such that the appropriate core operating limits for the applicable cycle will apply.

Technical Specifications will not be changed.

Sionificant Hazards Consideration NNECO has reviewed the proposed changes in accordance with 10CFR50.92 and has concluded that the changes do not involve a significant hazards consideration.

The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes would not involve a significant hazards consideration because the changes would not: '

l. Involve a significant increase in the probability or consequences of an accident previously analyzed.

lhe removal of cycle-specific core ope. rating limits from Technical Specifications has no influence or impact on the probability or consequences of a Design Basis Accident. The cycle-specific core operating l iniits, although not in Technical Specifications, will be followed in the operation of Millstone Unit No. 3. The proposed ameninent still requires exactly 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 FSAR will be examined with respect to changes in cycle-dependent param'eters, 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 requirements of 10 CFR 50.59, ensures that future reloads will not involve a significant increase in the probability or consequences of an g

accident previously evaluated.

2. Create the possibility of a new or different kind of accident from that previously analyzed.

As stated earlier, the removal of the cycle-specific variables 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 this pro-posed change. The cycle-specific variables are calculated using the

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U.S. Nuclear Regulatory Commission B13435/Page 4 February 21, 1990 NRC-approved methods and submitted to the NRC to allow the Staff to continue to trend the values of these limits. The Technical Specifica-tions will continue to require operation within the required core operat-ing limits and appropriate actions will be taken when or if limits are exceeded.

Therefore the proposed amendment does not in any way create the possibil-ity of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The margin of safety is not affected by the removal of cycle specific core operating limits from the Technical Specifications. The margin of safety presently provided by current Technical Specifications remains unchanged. Appropriate measures exist to control the values of these cycle-specific limits. The proposed amendment continues to require operation within the core limits as obtained from the NRC-approved reload design methodologies and appropriate actions to be taken when or if limits are violated remain uncnanged.

The development of the limits for future reloads will continue to conform to those methods described in NRC-approved documentation. In addition, each future reload will involve a 10 CFR 50.59 safety review to assure that operation of the unit within the cycle specific limits will not l involve a significant reduction in a margin of safety.

1 Therefore, the proposta changes are adninistrative in nature and do not impact tho operation of Millstone Unit No. 3 in a manner that involves a reduction in the margin of safety.

L Moreover, the Commission has provided guidance concerning the application of l standards in 10CFR50.92 by providing certain examples (March 6, 1986, l- SlFR7751) of amendmor.ts that are considered not likely to involve a significaat hazards consideration. Although the proposed changes are not enveloped by a specific example, the proposed changes would not involve a significant increase in the probability or consequences of an accident previously analyzed. The removal of cycle dependent variables from the L Technical Specifications has no impact upon plant operation or safety. Since g -

the applicable FSAR limits will be maintained and the Technical Specifications

'will continue to require operation with the core operating limits calculated l by the NRC-approved methodology, the changes do not reduce the effectiveness L of Technical Specification requirements. In addition, the action statements l and surveillance requirements will remain in the Technical Specifications.

Any changes to the COLR will be made in accordance with the provisinns of 10CFR50.59.

Based upon the information contained in this submittal and the environmental assessment for Millstone Unit No. 3, there are no radiological or l

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l l-t U.S. Nuclear Regulatory Commission B13435/Page 5 February 21, 1990 L

nonradiological impacts associated with the proposed change and the proposed license amendment will not have a significant effect on the quality of the human environment.

The Millstone Unit No. 3 Nuclear Review Board has reviewed and approved the attached preposed revisions and has concurred with the above determinations.

The proposed changes need to be approved to support Cycle 4 operation and prior to entry into Mode 4 after the November 1990 refueling outage. NNECO requests that these proposed changes be approved and effective prior to the startup of the ' upcoming refueling outage, which is presently scheduled for November 17, 1990. This would allow specific applicable Technical Specifi-cations to be in place prior to any affected mode change.

In accordance with 10CFR50.91(b), we are providing the ' State of Connecticut with a copy of this proposed amendment. ,

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Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Mv&

E. K firoc'zka SenioFVicePres/~tiden  ;

cc: W. T. Russell, Region I Administrator D. H. Jaffe, NRC Project Manager, Millt. tone Unit No. 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 Hr. Kevin McCarthy .

Director, Radiation Control Unit Department of Environmer,tal Protection Hartford, Connectic1t 00116 STATE OF CONNECTICUT)

COUNTY OF HARTFORD 1ss. Berlin Then personally appeared before me, E. J. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensee herein, and that the statements contained in said information a e true and correct to the best of his knowledge and belief.

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t Attachment Millstone Unit No. 1 Description of the Proposed Technical Specification  !

Changes - Cycle-Specific Parameter Linits l

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February 1990 1

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Millstone Unit No. 3 ,

Description of the Proposed Technical Specification  :

Chanaes - Cycle-Soecific Parameter Limits i

htgg Technical Soecification Chanae Description 1-7 1.42 Definitions Adds definition of COLR ,

3/4 1 4 3.1.1.3 Hoderator Temperature Relocates MTC Limits to the I Coefficient COLR 3/4 1-5 4.1.1.3 MTC Surveillance  :

Requirement B3/4 1-2 3/4.1.1.3 Moderator Temperature '

Coefficient Bases I

3/4 1-14 and 3.1.3.1 Moveable Control Replaces reference to Figures 3/4 1-15 Assemblies 3.1-1 and 3.1-2 (which are relocatedtotheCOLR)with Specification 3.1.3.6 3/4 1-20 3.1.3.5 Shutdewn Rod Insertion Replaces the fully withdrawn Limit limit with more general 4.1.3.5 Surveillance Requirement insertion limits which are to be provided in the COLR. This it done to allow RCCA repositioning to minimize wear.

l 3/4 1-21 3.1.3.6 Control Red Insertion Relocates the control bank rod ~

J Limits insertion limits to the COLR.

3/4 1-22 Figitre 3.1-1 Rod Banr. Insertion Relocate the figures to the COLR l Limits vs. Thermal Power, Four loop Operation 3/4 1-23 Figure 3.1-2 Rod Bank Insbrtion Limits vs. Thor:nal Power, Three Loop Operation 3/4 2-1 3.2.1.1 Axial Flux Difference Relocates the AFD Limits to the 3/4 2-2 (fourLoopsOperating)' COLR.

g 3/4 2-3 Fig. 3.2-la Axial Flux Difference Limits As a function of Rated Thermal Power (Four Loops Operating) 3/4 2-4 3.2.1.2 Axial Flux Difference 3/4 2 (Three Loops Operating)

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Attachment I  ;

! 'o 'B13435/Page 2  ;

Eitgg Technical SDecification Chance Descriotion -

3/4 2-6 Fig. 3.2-lb Axial Flux ,

Difference Limits As a function '

of Rated Thermal Power (ThreeLoopsOperating)

B 3/4 2-1 and 3/4.2.1 Axial Flux Difference B 3/4 2-2 l 3/4 2-7 3.2.2.1 Heat Flux Hot Channel Relocates the FQ, K(z) functions  ;

Factor (FourLoopsOperating) to the COLR. The numerical FQ limit.isrep1ggdwitha i 3/4 2-8 4.2.2.1.2.c, e, Heat flux function of F which is to be and specifiedin9heCOLR.

3/4 2-9 Hot Channel Factor Surveillance Requirements.  ;

3/4 2-10 Fig. 3.2-2a K(z) Normalized Fg (z) (for Four Loop Operation)

I 3/4 2-11 3.2.2.2 Heat Flux Hot Channel Factor (Three Loops Operating) 3/4 2-13 4.2.2.2.3, Heat Flux Hot Channel Factor Surveillance Requirements 3/4.2-14 Fig. 3.2-2b K(z) Normalized F (z) as a function cf tore h0ight. (For Three loop Operation).

B 3/4 2 6 3/4 2.3 Heat 519x Het Channal .

Factor 3/4 2-IS 3/4.2.3 RCS Flow Rate and Relocate the F-delta H limits to Nuclear Enthalpy Rise Hot the COLR. The numerical value Channel Factor - Four loop:: fortheF-deltagT)sreplared Operating with parameter F and PF '

which are definebHintheCb[4R. '

3/4 2-18 3/4.2.3.1 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor - Three Loops Operating, b

6 21 6.9.1.6 Radial Peaking Factor The Radial Peaking Factor Limit.

Limit Report Report is replaced with the description of the COLR. The COLR will be provided, upon issuance of each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrative and Resident Inspector.

In addition, the Technical Specification index has been revised to reflect the above changes.

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