ML20210D324

TS Proposed Change 218 to License DPR-28,evaluating Increase in Allowable Core Flow from 100% to 107%.Listed Specific Change Proposed

ML20210D324
Person / Time
Site:	Vermont Yankee
Issue date:	07/20/1999
From:	Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20210D326	List: ML20210D324 ML20210D337
References
BVY-99-82, NUDOCS 9907270197
Download: ML20210D324 (9)

Text

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'i-f VERMONT YANKEE l

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NUCLEAR Pow'ER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257 5271 July 20,1999 BW 99-82 U.S. Nuclear Regulatory Commission ATfN: Document Control Desk Washington, DC 20555

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specification Proposed Change No. 218 Enhancements to Support Imnlementation ofIncreased Core Flow Pursuant to 10CFR50.90, Vermont Yankee (W) hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed change into the W Technical Specifications.

W is evaluating an increase in allowable core flow from 100% (48.00 Mlb/hr) to 107% (51.36 Mlb/hr).

As a part of this evaluation, Technical Specifications and Bases enhancements were identified that are warranted such that the Technical Specifications and Bases reflect the plant configuration following implementation ofincreased core flow. The following specific change is proposed:

A high " clamped" value is added to the Table 3.2.5, Rod Block Monitor Upscale (Flow Bias) Trip Setting equation, e

A high " clamped" value is added to the Table 3.1.1, APRM High Flux (Flow Bias) equation for the reactor protection system trip, e Changes are made to clarify Figure 2.1-1, APRM Flow Reference Scram and APRM Rod Block Settings, and e The Bases for the affected sections are revised.

Attachment I to this letter contains supporting information and the safety assessment of the proposed change. Attachment 2 contains the determination of no significant hazards consideration. Attachment 3 provides the marked-up version of the current Technical Specification and Bases pages. Attachment 4 is the retyped Technical Specification and Bases pages.

W has reviewed the proposed Technical Specification and Bases change in accordance with 10CFR50.92 and concludes that the proposed change does not involve a significant hazards consideration.

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VY has also determined that the proposed change satisfies the criteria for a categorical exclusion in accordance with 10CFR51.22(c)(9) and does not require an environmental review. Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or cnvironmental assessment needs to be prepared for this change.

If you have any questions on this transmittal, please contact Mr. Thomas B. Silko at (802) 258-4146.

Sincerely, VE YANKEE NUCLEAR POWER CORPORATION An Robert J. Wan k //#

Director of Sa and Regulatory Affai _.

p d ,, SAtt x STATE OF VERMONT )

)SS WINDHAM COUNTY NOTAR) k

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'Iben personally appeared before me, Robert J. Wanczyk, who, being duly sworn, di th e is Dir ~ or p Safety and Regulatory Affairs of Vermont Yankee Nuclear Power Corporation, that he i utho ex /

and file the foregoing document in the name and on the behalf of Vermont Yankee Nucle that the statements therein are true to the best of his knowledge and belief.

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$0D Sally A. Saddstrum, Notary Public My Commission Expires February 10,2003 Attachments cc: USNRC Region 1 Administrator USNRC Resident Inspector- VYNPS USNRC Project Manager- VYNPS Vermont Department of Public Service l

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VEnuoNT hun Nn:uan Powru Com tmuios

- 1 Docket No. 50-271 l

BVY 99-82 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 218 Enhancements to Support Implementation ofIncreased Core Flow Supporting Information and Safety Assessment of Proposed Change

"""" ""# "" """" C""Po" An o N BVY 99-82 / Attachment I / Page i INTRODUCTION Vermont Yankee (VY) has Reactor Recirculation System capacity in excess of that required to provide rated core flow (48 Mlb/hr). Allowing core flow to exceed rated core flow would provide greater flexibility in attaining and maintaining rated power. This flexibility could be used during startups and control rod pattern adjustments to compensate for the effects of xenon and fuel burnup without adjusting control rods. At end of cycle, recirculation flow could be increased above 100% rated flow to allow longer continued operation at 100% power. Therefore, VY is currently evaluating the feasibility of increasing core flow.

Technical Specifications and Bases changes have been identified that are warranted such that the Technical Specifications and Bases reflect the plant configuration following implementation of increased core flow. Moreover, these changes are applicable and desirable regardless of the result of the increased core flow feasibility evaluation. Each proposed change and its basis are detailed below. Of these changes, only those that relate to clamping the maximum value of the Rod Block Monitor Upscale (Flow Bias) function are directly necessitated by increased core flow. However, those changes simply ensure transient analysis assumptions are maintained and neither provide relief nor add restrictions to current operating conditions and regulatory requirements. l

1. For Figure 2.1-1, APRM Flow Reference Scram and APRM Rod Block Settings, the clarifying statement "Setpoints shall be $ values shown on the graph" is proposed to be added.

Basis: Technical Specifications Section 2.1 states, for both the APRM Flux Scram Trip Setting and the APRM Rod Block Trip Setting, that the " setting shall be as shown on Figure 2.1.1." Since Figure 2.1.1 only shows straight line graphs of recirculation flow versus neutron flux, the Section 2.1 statements can be restrictively interpreted to mean that the setpoint must be an exact point on the line. 'Ihe lines actually are the graphical representation of the scram and rod block equations. The trip setpoints must be less than or equal to the value represented by the line.

This change is considered administrative in that it adds clarification to Figure 2.1.1 that already exists in Technical Specifications Section 2.1.

2. For Bases Section 2.1.B, page 16, and for the Bases Section 3.2 APRM rod block trip discussion, l page 77, the current Bases is proposed to be replaced with a more accurate discussion of the function, as identified in the VY Final Safety Analysis Report (FSAR). l Basis: The current discussion of the APRM Rod Block Trip Setting is ambiguous. It implies that the function is assumed in the protection of the fuel cladding integrity safety limit. In fact, the APRM  ;

Rod Block Trip Setting is not assumed in any accident or transient analysis. It is provided to stop 1 reactivity addition prior to reaching the scram setpoint. This is fully discussed in the FSAR Sections 7.7.4.3.2 and 7.5.7.3.

This change is considered administrative in that it clarifies the Bases, providing information taken verbatim from the FSAR.

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3. For Table 3.1.1, Reactor Protection System (Scram) Instrument Requirements, Average Power Range Monitor Upscale (Flow Bias) function, it is proposed to add "with a maximum of 120%" to the APRM High Flux (Flow Bias) Trip Function equation.

Basis: This change is consistent with Specification 2.1.A.I.a, which states: "For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be I

allowed to exceed 120% of rated thermal power." The limitation becomes significant with increased core flow as the trip setting equation yields a value that exceeds 120% for core flow values in excess of 100%.

l This change is considered administrative in that it adds information to Table 3.1.1 that already exists

! in Technical Specifications Section 2.1.

4. For Table 3.2.5, Control Rod Block Instrumentation, Rod Block Monitor Upscale (Flow Bias) function, the caveat "with a maximum a.; defined in the COLR" is added to the Trip Setting equation.

Basis: This function is assumed in the mitigation of the Continuous Rod Withdrawal During Power l

Range Operation transient. The trip setting is defined by Technical Specifications as l

< 0.66(W-AW) + N, where W is percent rated two loop drive flow, AW is the difference between the two loop and single loop drive flow at the same core flow and must be accounted for during single loop operation, and N is a variable defined in the Core Operating Limits Report (COLR). Inputting the increased core flow maximum flow value of 107% would raise the trip setting above the analyzed limit. To ensure the assumption of the transient analysis is met, the trip setting value must be l clamped to not exceed the valu: yielded by the equation when 100% core flow is input. Since the "N" value in the equation is cycle specific and is a variable defined in the COLR, the clamped value will also be cycle specific and defined in the COLR.

This change is more restrictive than the current Technical Specifications in that it provides an upper limit on the equation. As such, the assumptions of the Continuous Rod Withdrawal During Power Range Operation transient are ensured.

5. For Bases page 77, it is proposed to delete the current paragraph describing the control rod block systems and replace it with the following: "The trip logic for the nuclear instrumentation control rod block logic is I out of n; i.e., any trip on one of the six APRMs, six 1RMs or four SRMs will result in a rod block. The minimum instrument channel requirements for the IRM may be reduced by one for a short period of time to allow for maintenance, testing or calibration. The RBM is credited in the Continuous Rod Withdrawal During Power Range Operation transient for preventing excessive control rod withdrawal before the fuel cladding integrity safety limit (MCPR) or the fuel rod mechanical overpower limits are exceeded. The RBM upper limit is clamped to provide protection at greater than 100% rated core flow. The clamped value is cycle specific; therefore, it is located in the Core Operating Limits Report."

Basis: The current paragraph is ambiguous in its portrayal of the safety significance of the control rod block functions. ; The new paragraph retains the discussions of the trip logic and IRM maintenance / testing, clarifies the ambiguities regarding the purpose of the nuclear instrumentation rod blocks, and adds discussions of the purpose of the RBM and the clamped upper limit. The purpose of the RBM discussion is taken verbatim from the FSAR Section 14.5.3.1, Continuous Rod Withdrawal During Power Range Operation.

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• This ' change is considered administrative in that it clarifies information currently in the Bases, l provides additional discussion of the RBM upscale clamp, and adds a discussion of the purpose of l the RBM taken directly from the FSAR. l Note that due to format enhancements and presentation consistency, Bases pages 16,17, and 75 through 79 are retyped and provided in Attachment 4 of this document. However, the changes proposed herein l are found specifically on Bases pages 16, 77, and 78. Bases page 75 also contains an editorial change i separating a note from the body of the discussion.

SAFETY ASSESSMENT The proposed change is not a safety concern and can be implemented without endangering the public health and safety becaase:

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1. Changes 1 and 3 are administrative and only clarify specifications by adding information that is already specified elsewhere in Technical Specifications. These changes have no impact on technical content.

2. Changes 2 and 5 clarify ambiguities in the Bases, using wording taken directly from the FSAR.

Discussion is also added describing the clamped upper value on the Rod Block Monitor Upscale (Flow Bias) trip. The wording is descriptive only and does not change the meaning or intent of the specification.

3. Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum value requirement to the Technical Specifications. Limiting the upscale trip setting at flows in excess of 100% of rated core flow ensures the assumptions of the Continuous Rod Withdrawal During Power Range Operation transient are met.

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Docket No. 50-271 i BVY 99-82 1

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l Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 218

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Enhancements to Support Implementation ofIncreased Core Flow Determination of No Significant Hazards Consideration I

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Porsuant to 10CFR50.92, Vermont Yankee (VY) has reviewed the proposed changes and concludes that the changes do not involve a significant hazards consideration since the proposed changes satisfy the criterla in 10CFR50.92(c).

1. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a sienificant increase in the probability or consecuences of an accident oreviously evaluated.

Changes 1 and 3 are administrative and have no impact on technical content; therefore, they do not increase the probability or consequences of an accident previously evaluated.

Changes 2 and 5 clarify ambiguities in the Bases. The wording is descriptive only and does not change the meaning or intent of the specification. Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum value limitation to the Technical l Specifications. Limiting the upscale trip setting at flows in excess of 100% of rated core flow ensures l the assumptions of the Continuous Rod Withdrawal During Power Range Operation Transient are met.

No other accident or transient analyses are affected. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in accordance with the cronosed amendment will not create the possibility of a new or different kind of accident from any accident oreviously evaluated.

Change 4, limiting the maximum value for the Rod Block Monitor Upscale (Flow Bias) function, is a change to plant design, in that it clamps the upscale trip setting at flows in excess of 100% of rated core flow at the 100% core flow value. This change ensures the assumptions of the Continuous Rod .

Withdrawal During Power Range Operation Transient are met and has no effect on any other accident or transient analyses. Changes 1,2,3, and 5 do not involve a change to the plant design.

None of the proposed changes affects any parameters or conditions that could contribute to the initiation of any accident. No new accident modes are created. No safety-related equipment or safety functions, other than the Rod Block Monitor as discussed above, are altered as a result of these changes.

Based on the above VY has concluded that the proposed change will net create the possibility of a new or different kind of accident from those previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a sienificant reduction in a marcin of safety.

Changes I and 3 are administrative and have no impact on technical content. Therefore, they have no effect on margin of safety.

Changes 2 and 5 clarify ambiguities in the Bases, using wording taken directly from the FSAR. The wording is descriptive only and does not change the meaning or intent of the specification. Therefore, these changes do not involve a significant reduction in a margin of safety.

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BVY 99-82 / Attachment 2 / PCge 2 Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum value limitation to the Technical Specifications. Limiting the upscale trip setting at flows in excess of 100% of rated core flow ensures the assumptions and, therefore the margin of safety, of the Continuous Rod Withdrawal During Power Range Operation transient are met. No other accident or transient analyses are affected. Therefore, this change does not involve a significant reduction in a margin of safety.

Summary No Sienificant Hazards Consideration On the basis of the above, VY has determined that operation of the facility in accordance with the proposed changes does not invnive a significant hazards consideration as defined in 10CFR50.92(c), in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

In making this determination, Vermont Yankee has also reviewed the NRC examples of license amendments considered not likely to involve significant hazards considerations as provided in the final adoption of 10CFR50.92 published in the Federal Register. Volume 51, No. 44, dated March 6,1986.

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