ML20116A853

Application for Amend to License NPF-42,revising TS to Support Use of Vantage 5H Fuel W/Intermediate Flow Mixers Beginning in Cycle 7 to Include Results of Analyses Performed for Operation at Increased Power Level

ML20116A853
Person / Time
Site:	Wolf Creek
Issue date:	10/28/1992
From:	Hagan R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20116A856	List: ML20116A853 ML20116A859 ML20116A868 ML20116A873
References
NA-92-0073, NA-92-73, NUDOCS 9210300184
Download: ML20116A853 (11)

Text

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WOLF CREEK ' NUCLEAR OPERATING CORPORATION =

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SN. MEff" or Aou,anc. October 28, 1992-NA 92-0073 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station F1-1.7 Washington, D. C. 20555  ;

Subject:

Docket No. 50-482: Revision to Technical Specifications-for Cycle 7 Gentlemen E

-This letter transmits an application for amendment to Facility Operating i License No. NPF-42 for Wolf Creek Ger.erating Station (WCGS), . Unit 1. This lit. ens e ameadment request proposes revisions to various Technical Spec 3fication_ sections to support the use of VANTAGE SH fuel with Intermediate

-Flow Mixers beginning in cycle 7; to include results of analyses performed for o: era tion at-an increased power level; to allow main steam safety valve setpoint tolerance increases: and to relocate cycle-specific parameters to the Core Operating Limits Report.

The analyser performed to support these revisions were done in accordance with the .- me thodolohies described in Wolf Creek Nt.clea r Operating Corporation's (WCNOC) topical- reports shich are currently with the NRC for approval.

Although the analyses have been performed for operation at an increased power level, approval for_ operation at this level is not being requested at this time. The - relocation of cycle-specific parameters to the Crr. Operating Limi- Report is - being done as all iwed by Generic Letter 88-16 " Removal-of Cyci specific Parameter Limits frut Technical Specifications."

Implementation of this proposed revisica to the WCGS technical specifications ,

is needed to support the upcoming refueling outage. -Therefore, WCNOC requests. i approval of this proposed amendment prior to.the sixth refueling. outage which is scheduled-to begin in March, .1993. The proposed revision will be fully implemented within 30 days of formal Nuclear Regulatory Commission appreval.

Attachment I,-provides a; safety evaluation, Attachment II provide _s a significant hazards _ consideration determination, . and Attachment III provides an environmental impact determination. Attachment IV ' provides the specific-changes to the . technical specification and the detailed evaluation / analyses associated with each change. .

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1 Dox a n / Burlington, KS 66839 / Phone: (316) 364-8831 -

9210300184 921028 2 An Eaual oppoqunity Employer M/F/HCWET

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- Page 2 of 2 In accordance with 10 - CFR 50.91, a copy of this application, with attachments, is being provided'to the designated Kansas State Official.

If you have any questions concerning this matter, please contact me at (316) 364-8831 Ext. 4553 or Mr. Kevin J. Moles of my staff at (316) 364-8831

-- Ext. 4565.

Very Atruly yours, y-

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i R$bert C. Hagan g,'

Vice President Nuclear Assurance RCH/jra

. Attachments: I - Safety Evaluation II - Significant Hazards Consideration Determination III - Environmental Impact Determitation IV - Safety Evaluation and Cycle 7 Technirsl Specification-Change Report

- cc -G'. W Allen (KDHE), w/a A'. T.-Howell (NRC), w/a J. L. Millican (NRC), w/a G. A. Pick (NRC), w/a W. - D. Reckley'(NRC), w/a

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1 STATE OF KANSAS }

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1 Robert C. Hagan, of lawful age, being first duly sworn upon oath says that he is Vice President Nuclear Assurance of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of said Corporation with full power and authority to dc so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

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By 8 1 h/L / !fM [ _

Rob"erjtC.'Hagan /

Vice Uresident ,#

I Nuclear Assurance SUBSCRIBED and sworn to before me this c2E day of CCl , 1992. ,,,.." "*"<,,,

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Attachment I to NA 92-0073 Page 1 of 2

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ATTACIIMENT I SAFETY EVALUATION

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Attachment I to NA 92-0073 Page 2 of 2 Safety Evaluation

!- Proposed Channe The proposed technical specification revisions are being made to support the .

use of VANTAGE SH fuel with the Intermediate Flow Mixer (IFM) grid feature incorporated. In addition this evaluation supports the analyses associated with operation at an increased power level, an increase in main steam safety valve setpoint tolerances, and the relocation of_ cycle-specific parameters to

,- the Core Operating Limits Report, j Evaluation The changes proposed above were evaluated using the following criteria to determine any effects on safe operation of the plant.

1. Departure from Nuclear Boiling (DNB) design basis

2. Fuel Temperature Design Basis

3. Reactor Coolant System Pressure

4. Loss of Coolant Design Basis The evaluation considered non-LOCA transients. LOCA transients, and the radiological consequences of accidents. The results showed that; the proposed changes and assumptions are not initiators of any accidents the consequences of evaluated accidents are not increased and the radiological consequences remain well within 10 CFR 100 guidelines _for a full core of. VANTAGE 5H with IFM as well as the transition cores; and no new performance requirements are '

being imposed on any system or component to _ _ support ~ the revised analysis.

l Therefore, there is no change in the probability or consequences of the types_

( of accidents previously evaluated._ __ The evaluation also _ showed that~ with_ the

!_ proposed changes and assumptions the operating envelope'would-remain bound by, and in -no ca r.e exceed the acceptance limits defined in the technical' specifications.

L . Based en the - above findings and the detailed evaluations included with l Attachment IV, the proposed changes - do not increase fthe. probability of occurrence or the consequences of an accident - or malfunction of equipment important to . .saf ety previously ~ evaluated in the safety . analysis , report, or create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis reports _or reduce the margin ~

of safety as defined.in the basis for any technical. specification. _Therefore, the proposed r$anges do-not adversely affect or endanger.the health or safety of the general public or involve a significant _ safety hazaru.

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Attachment 11 to NA 92-0073 Page 1 of 2 I

ATTACIDiENT II SIGNIFICANT lit.ZARDS CONSIDERATION DETERMINATION C

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Attachment-II to NA 92 0073 Fage 2 of 2 Significant 11azards Consideration Determination The proposed technical specification revisions are being made to support the use of VANTAGE SH fuel with Intermediate Flow 7"xers beginning in cycle 7; to

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include results of analyses performed for an increased power level _ an increase in main steam safety vcive setpoint tolerancest and to relocate cycle specific parameters to the Core Operating Limits Report as allowed by Generic Letter 88-16.

The following standards were considered in tr 4 mining if any significant hazards would be created by this proposed change.

Standard 1 - Involve a Sinnificant Increase in the Probability or Conneouences of an Accident Previously Evaluated The proposed changes have been found not to be initiators of any accident and therefore do not affect the probability of occurrence. The proposed changes '

do not introduce any new performance requirements for any -systems or components, therefore, the ability of the_ systems or-components to limit the consequences of an accident is not decreased.

Standard 2 - Create the Possibility of a New or Different Kind of Accident From any Previousiv Evaluated The proposed changes do not introduce any new performance requirements or modes of operation to any systems or components and therefore, do not introduce any new f ailure modes. Therefore, there is no possibility of the-creation of a new or different kind of accident from any previously evaluatsd.

Standard 3 - Involve a-Sinnificant Reduction in the Marnin of Safety An evaluation of the proposed changes and assumptions ha's shown that operation under the new conditions would remain bound by the: operating envelope' defined by technical specifications. Therefore, the margin of safety is not' reduced.

Based' on the above discussion it has been de te rmined .: that the requested technical specification revision does not violate any of the three standards.

Therefore, the requested license - amendment does not - involve a significant.

hazards consideration.

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Attachment III to NA- 92-0073-Page 1 of 3 ATTACllMENT III ENVIRONMENTAL IMPACT DETFRMINATION p-i

Attachment III to NA 92-0073 Page 2 of 3 l l

Environmental ]mpact Determination This amendment request meets the criteria specified in 10 CFR 51.22(c)(9).

Specific criteria contained in this section are discussed below.

(1) this amendment involves no significant hazards consideration As demonstrated in Attachment II. this proposed amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite this proposed technical specification change involves the following F

• Incorporating Intermediate Flow Mixer grid feature of the VANTAGE 5H fuel assembly Increase in Fag /Fq e Positive Moderator Temperature Coefficient

• OTAT/0 PAT Setpoint Changes

• Decrease in Allowed RCS Thermal Design Flowrate

• Main Steam Safety Valve Setpoint Tolerance Increase

• Shutdown Margin Increase in Mode 5

• Core Operating Limit Report In addition to the technical specification changes noted above, the following ,

conservative assumptions were incorporated into the analysis:

• Increase rated thermal power (from 3411 MWt to 3565 MWt) '

• Analysis over a range of temperatures

• 10% steam generator tube plugging

• -Thimble plug-removal The current radiation source terms used in the current Updated Safety Analysis Report-(USAR) dose consequences analyses were based on a core power level of 3565 MWt and on a 12 month operating cycle. The radiation source terms were then re-calculated when' the Jesign change that converted the 12-month fuel cycle to 18-month was implemented. The impact of extended fuel burnup-as a result of increased cycle length on the radiological consequences of-accidents was found to be insignificant since the calculated radiation - doses are directly related to the core inventory of radioactive isotopes which is not strongly influenced by the extension of fuel burnup.

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.__._>..m_..._.m Attachment III to NA 92-0073 Page 3 of 3 l

l While the effects f rom the t nsition 9 an extended fuel cycle and a core thermal power of 3565 IEt on these s.nalyses have been evaluated to be minor, Wolf Creek . Nuclear Operating Corporation (WCNOC) has elected to use the revised source terms to update the radiological consequ'.nce analyses contained in the USAR. The revised source terms are describA for the various USAR Chapter 15 accidents in Section 5.7 of Attachment IV.- The revised source terms will not re: ult in any eignificant chang in the types or sign 3ficant increase in the amounts of any effluents that may be released offsite.

(iii) there is no s!.gnificant increase in individual or cumulative-occupational radiation exposure As described above, the analysis shows the revised source term changes to be '

minor. Therefore, there will be no significant increase in individual or cu'nul a t ive occupational radiation exposure associated with the proposed change.

Based on the above, there will be no significant impact 7n the environment

- re su1 *.ing from this change and the change meets the criteria specified in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements of 10 CFR 51.21 relative to a specific environmental assessment by. the Commission, i

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Att achtnent IV to . NA 92-00_7 3

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-l ATTACIMENT IV i

i Safety Evaluation and Cycle 7 Technical Specification Change Report I

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