ML20209G167

Proposed Tech Specs Revising Value for SLMCPR & Deleting Wording Which Specifies These as Cycle 20 Values

ML20209G167
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	07/12/1999
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20137X301	List: ML20209G163 ML20209G167
References
NUDOCS 9907190064
Download: ML20209G167 (9)

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Docket No. 50-271 i

BVY 99-91 l

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Attachment 3 l

Vermont Yankee Nuclear Power Station

! Proposed Technical Specification Change No. 221 l

l Safety Limit - Minimum Critical Power Ratio Marked-up Version of the Current Technical Specifications i

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i l-9907190064 990712 PDR ADOCK 05000271 P PDR  !

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTFGRITY 2.1 FUEL CU ODING INTEGRITY Apolicability: Aeolicability:

Applies to the interrelated Applies to trip setting of the l variable associated with fue', .

instruments and devices which therral behavior, are provided to prevent the nuclear system safety limits from being exceeded.

Obiective Obiective:

To establish limits below which To define the level of the the integrity of the fuel process variable at which cladding is preserved. automatic protective action is initiated.

Specification: Specification:

A. Bundle Safety Limit (Reactor A. Trio Settinas Pressure >000 osia and Core Flow >10% of Rated) The limiting safety system trip settings shall be as When the reactor pressure is specified below:

>800 psia and the core flow is greater than 10% of 1. Neutron Flux Trio rated: Settines l 1. } e/Cy le/20 o a. APRM Flux Scram Trio ind. he /ex t* Settino (Run Mode)

A Minimum CritiJ,1 Power Rati CPR) L / /2 When the mode switch l less than, ( for is in the RUN Single L6o era lon) position, the APRM l shall constitute flux scram trip i f./0 violation of the Fuel setting shall be as I Cladding Integrity shcwn on l Safety Limit (FCISL). Figure 2.1.1 and i FC e o i s 1 shall be:

l t el 2 w ub 1 ee @i I ec le a 'o of the S<0.66(W-AW)+54%

t C .

i l where: i S = setting in percent of l rated l thermal power j (1593 MWt)

W = percent rated two loop drive ,

flow where '

100% rated drive flow is that flow equivalent

'~ to 48 x 10 8 lbs/hr core flow Amendment No. M , 44, 64, M, 94, M9, MG, 159 6

vuoioxr vmu>. Necu.sn Powr.n coni oiuuos Docket No. 50-271 BW 99-91 l

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l l Attachment 4 l

l Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 221 1

Safety Limit - Minimum Cri,tical Power Ratio l Retyped Technical Specification Pages i

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VYNPS 1.1 , SAFETY' LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING l 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which are thermal behavior, provided to prevent the nuclear system safety limits from being exceeded.

Objective: Objective:

'To establish limits below which To define tbs level of the process the integrity of the fuel cladding variable at which automatic is preserved. protective action is initiated.

Specification:

Specification: -

A. Trip Settings  ;

A. Bundle Safety Limit (Reactor Pressure >800 psia and Core The limiting safety system Flow >10% of Rated) trip settings shall be~as i specified below:

When the reactor pressure is

>800 psia and the core flow is 1. Neutron Flux Trip Settings i greater than 10% of rated: i

a. APRM Flux Scram Trip l

1. A Minimum Critical Power Setting (Run Mode) l Ratio'(MCPR) of less than 1.10 (1.12 for Single Loop When the mode switch

-l Operation) shall constitute is in the RUN violation of the Fuel ' position, the APRM Cladding Integrity Safety flux scram trip Limit (FCISL). setting shall be as shown on Figure 2.1.1 l and shall be:

S50.66(W-AW)+54%

where:

S = setting in percent of rated thermal power (1593 MWt)

W = percent rated two loop drive flow i where 100% )

rated drive i flow is that flow equivalent to 48 x 10' lbs/hr core i flow Amendment ' No. M, M, 44, M, 44, M9, M9, M9 6

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vnuion nsun secuan vown< coiwouxuos Docket No. 50-271 BW 99-91 Attachment 6 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 221 General Electric Summary of Technical Basis for SLMCPF, /elues (Non-Proprietary Version) i i

Attaciamtnt Additi:n:1Inf:rmiti:n Reg:rding the June 14,1999 Cycle Specific SLMCPR for Vermont Yankee Cycle 21 References

[1] Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE)," Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P. Methodology and Uncertaintiesfor Safety Limit 1 MCPR Evaluations; NEDC-32694P, Power Distribution Uncertaintie.rfor Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle Specific Safety Limit MCPR,"

(TAC Nos. M97490, M99069 and M97491), March 11,1999.

[2] Letter, Thomas H. Essig (NRC) to Glen A. Watford (GE)," Acceptance for Referencing of Licensing Topical Report NEDC-32505P, Revision 1, R-Factor Cakulation Methodfor Gell.

GE12 and GE13 Fuel,"(TAC No. M99070 and M95081), January 11,1999.

[3) General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.

Comparison of Vermont Yankee Cycle 21 and Cycle 20 SLMCPR Values Table I summarizes the relevant input parameters and results of the SLMCPR determination for the Vermont Yankee Cycle 21 and Cycle 20 cores. The SLMCPR evaluations were performed using NRC approved methods and uncertainties [Il. These evaluations yield different calculated SLMCPR values because dif": rent inputs were used. The quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided.

In comparing the Vermont Yankee Cycle 21 and Cycle 20 SLMCPR values it is important to note the impact of the differences in the core and bundle designs. These differences are summarized in Table 1.

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11 The uncontrolled bundle pin-by-rh power distributions were compared between the Vermont Yankee Cycle 21 bundles and the Cycle 20 bundles. Pin-by-pin power distributions are ,

characterized in terms of R-factors using the NRC approved methodology [21, ((

11 Summary

((' )) have been used to compare quantities that impact the calculated SLMCPR value.

Based on these comparisons, the conclusion is imbed tLt* the Vermont Yankee Cycle 21 core / cycle has a more peaked core MCPR distribution (( ]) and flatter in-bundle power

(( GENE Proprietary Information )) page1of3

(( enclosed by double brackets ))

' Att chmnt Additi:n:1 Int:rm:ti:n R:g rdi:g the June 14,1999

, Cycle Specific SLMCPR for Vermont Yankee Cycle 21 i o

distributions (( ')) than what was used to perform the Cycle 20 SLMCPR ,

evaluation. '

The calculated 1.10 Monte Carlo SLMCPR for Vermont Yankee Cycle 21 is consistent with what i one would expect ((

)) the 1.10 SLMCPR value is appropriate.

Based on all of the facts, observations and arguments presented above, it is excluded that the calculated SLMCPR value of 1.10 for the Vermont Yankee Cycle 21 core is appropriate. Itis reasonable that this value is 0.01 lower than the 1.11 value calculated for the previous cycle.

For single loop operations (SLO) the calculated safety limit MCPR for the limiting case is 1.12 ((

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Prepared by: Verified by: l Db ^ - , ,

G.M. Baka G.N. Marrotte l Technical Project Manager Nuclear Fuel Engineering i Vermont Yankee Project i

(( GENE Proprietary Information )) page 2 of 3

{[ enclosed by double brackets ))

Attichment Additi:rlI:fsrm ti:o Regnrding the June 14,1999 Cycle Specific SLMCPR for Vermont Yankee Cycle 21 Tf.,ie1 Comparison of the Vermont Yankee Cycle 21 and Cycle 20 SLMCPR

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(( GENE Proprietary Information)) page 3 of 3

(( enclosed by double brackets )) ,

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Vr.nuoNT Ymut Necu:w Powut Colu onnuoN 1

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Docket No. 50-271 BW 99-91 i

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l Attachment 5 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 221 General Electric Summary of Technical Basis for SLMCPR Valuee.

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