ML20217F126

Forwards Update to Previously Submitted RELAP5 Analytical Assumptions for App R,Re RAI of 961104

ML20217F126
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	10/12/1999
From:	Leach D VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BVY-99-96, NUDOCS 9910200232
Download: ML20217F126 (9)

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VERMONT YANKEE l y NUCLEAR POWER CORPORATION "185 Old Ferry Road, Brattleboro, VT 05301 7002

• l (802) 257 5271 ]

October 12,1999 '

BVY 99-96 United States Nuclear Regulatory Commission l ATTN: Document Control Desk )

Washington,DC 20555

)

References:

(a) Letter, VYNPC to USNRC, " Response to Request for Additional Information l

Regarding 10CFR50, Appendix R Exemptions," BVY 96-139, dated November 4, (

1996.

(b) Letter, USNRC to VYNPC, " Request for Additional Information - V mont Yankee Request for Exemption From 10CFR50 Appendix R, Section 111.0, ' Fire Protection of Safe Shutdown Capabihty' and Section llI.L. ' Alternative and Dedicated Safe Shutdown Capability' (TAC Nos. M95149 and M95442)," NVY 96-146, dated September 20,1996.

(c) Letter, VYNPC to USNRC, " Request for Exemption from 10CFR Part 50, l Appendix R,Section III.G, ' Fire Protection for Safe Shutdown Capability' and {

Section III.L, ' Alternative and Dedicated Shutdown Capability'," BVY 96-67, i dated May 21,1996. I (d) Letter, USNRC to VYNPC, " Vermont Yankee Nuclear Power Station (TAC f Nos. M95442 and M95149)," NVY 97-128, dated August 12,1997. '

(e) NEDC-32814P, " Vermont Ycnkee Nuclear Power Station SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," March 1998.

(f) Vermont Yankee Nuclear Power htion Cycle 20, Core Operating Limits Report, Revision 3, September 1999.

Subject:

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

Update to Previousiv-Submitted RELAPS Analytical Assumptions for Appendix R In response to a Staff request contained in Reference (b), Vermont Yankee (VY) provided, in Reference (a), ,

the assumptions used during the RELAPS thermal-hydraulic analysis performed to support the safe  !

shutdown metho* (ADS /CS and ADS /LPCI) proposed in Reference (c). Following receipt of )

Reference (a), the NRC issued Refereece (d), which included a Safety Evaluation Report granting VY's j requested Appendix R exemptions. Subsequent to that time VY has explored the possibility of uprating l

. plant output from 100% power (presently 1593 MW th ) to 105% power (1673 MW th ), a move which, if )

completed, will require an update to the aforementioned RELAP5 analysis. In anticipation cf this

' development, VY has modified the previously-established assumptions. In addition, the assumptions have been adjusted to permit adoption of a cycle-independent analytical approach by selecting conservative values for decay heat, stored energy and power shapes, thereby alleviating the burden of performing a reanalysis prior to each operating cycle. Finally, VY is undergoing a gradual conversion to the GE-13 type fuel bundle in lieu of the GE-9 bundle used in the original analysis, affecting previous bundle power assumptions.

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fi . *. l y ., - q VERMONT YANXEE NUCLEAR POWER CORPORATION I~ BVY 99-% / Page 2 of 2 -

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.'The earlier assumptions affected by these changes have been undated to:

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e. Assume a 5% increase in reactor power to support a possible future power uprate. I l*. L Apply conservatism in the initial conditions to ensure cycle independence of the analysis.

.- Incorporate specific input changes in order to model the GE-13 fuel bundle.

e, incorporate input changes to assure consistency with the LOCA analysis of record (i.e. revised

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Safety Relief Valve setpoints).

'Because the subject assumptions may have been relied on by the Staff in formulating Reference (d), VY believes.it is appropriate to update the docket with this more recent information. Attached is a description of each assumption that was revised from the original contained in Reference (a), and the basis for its j acceptability.

Ifyou have any questions regarding this submittal, please contact Mr. Wayne M. Limberger at

-(802) 258-4237.

Sincerely, VERMONr A KEEN R POWER CORPORATION 0 l$ N"

.a onTL cF~ l Vice Pr i nt of Engineering Attachments cci . USNRC Region 1 Administrator USNRC Resident Inspector-VYNPS USNRC Project Manager-VYNPS i

. Vermont Department of Public Service b

Docket No. 50-271 BVY 99-96 l Attachment 1.

Vermont Yankee Nuclear Power Station Update to Previously-Submitted RELAPS  ;

Analytical Assumptions for Appendix R Revised Assumptions and Supporting Information l

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BVY 99-% / Auachment I / Page !

Previous Assumetion #1:

The reactor is assumed to be operating at full power, and at normal water levd (160 inches above top of 1 active fuel) at the time of event initiation.

Revised Assumotion-t The dactor is assumed to be operating in a power uprate (105%) condition and at normal water level inches above top of active fuel) at the time of event initiation. >

Justification:

De' higher power assumption is a conservative input when applied to the presently licensed 100% power level. It results in more stored energy and higher hot bundle power, which lead to a higher Peak Cladding

' Temperature (PCT). De assumption related to the normal water level of 160 inches above top of active fuel has not changed.

l Previous Assumption #6:

Decay heat is calculated by the 1979 ANS decay heat standard with 2-sigma uncertainty plus 7% power uncedainty (combined statistically) with a normal full-power of 1593 MW

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Revised Assumption:

Decay heat is calculated by the 1979 ANS decay heat standard with 2-sigre uncertainty plus 2% power

. uncedainty (combined statistically) with an uprated ful! power of 1673 MWth-Justification:

The decay heat is calculated using the same method described in Reference (a) but at a higher power level '

to allow for a 5% power uprate.

Previous Assumption #9:

Safety Relief Valve (SRV) setpoints assumed in the analysis are:  ;

Open (psjs!) Closed (psid)

SRVI 1080 1047.6

-. SRV 2/3 1090 1057.3 2 SRV 4 1100 1067.0 i

BVY 99-% / Attachment I / Page 2

Revised Assumotion': I i

.. Safety Relief Valve (SRV) setpoints assumed in the analysis are:

Ooen (osid) ' Closed (osid) -

SRV1 1113 1080.6

SRV 2/3 1123 1089.3' SRV 4 1133. 1099.0

. Justification:

The SRV setpoint assumptions were increased by 3% to allow for greater drift as permitted in the VY

- Technical Specifications, Table 2.2.1, Note 1. This is consistent with the assumptions used in

Reference (e). The change in setpoints has minimal effect on the Appendix R results. The change was made for consistency with the LOCA analysis of record and the Technical Specifications.

Previous Assumption #12:

Fuel stored energy in the hot bundle corresponds to a peak node power of 14.4 kW/ft plus fuel behavior code uncertainty, For the rest of the core, the fuel stored energy is calculated assuming a power level of 1593 M Wth-Revised Assumption:

The fuel stored energy in the hot bundle is calculated consistent with the power shape being analyzed and at the PLHGR consistent with that power shape. For the rest of the core, the fuel stored energy is calculated t

assuming a power level of 1707 Mwth (which allowr, for an additional 2% calorimetric uncertainty)._

l Justification:

I The stored energy in the core is overestimated for the present 100% power condition since the calculation is performed with 7% added power. This results in earlier core uncovery and, as a result, higher PCT. The l- change in the assumption for the hot bundle was made to assure consistency between the fuel performance and the RELAPSYA models. In the event power uprate is undertaken, this revised assumption will support

) an increase up to 105% power, after conservatively applying 2% additional calorimetric uncertainty as

noted in the revised assumption.

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f BVY 99-% / Attachment I / Page 3 i Previous Assumotion #13:

, Bounding axial power shapes were assumed based on plant data from Cycles 15 through 19. The power shapes used were the ones that produce maximum kW/ft seen at each axial level. The maximum average planar linear heat generation rate (MAPLHGR) was assumed to be 13.6 kW/ft which is bounding for the

. GE-9B bundle type presently used by Vermont Yankee. A variable hot bundle power was used depending i on the power shape with a maximum of 7.23 MWth which is much higher than the maximum bundle power experienced at the plant.

. Revised A'sumotion:

s Bounding axial power shapes are assumed based on plant data for Cycles 20 and 21. These power shapes are the ones that produce the maximum kW/ft seen at each axial level in the hot rod. Additional conservatism is introduced by assigning the power shapes derived from the plant data for the hot rod to the hot bundle. A variable hot bundle power will be used depending on the power shape with a maximum of

, 7.54 Mwth, which is much higher than the maximum bundle power experienced at the plant.  !

Justification; The changes in this assumption were needed to accommodate the'GE-13 bundles that were first loaded in

~the core for Cycle 20; hence, plant data are derived only for Cycles 20 and 21. The assumption that the hot

- bundle is at the axial power shape of the hot rod is very conservative. The increase in bundle power is appropriate for the GE-13 bundle, which can operate at a higher bundle power than the GE-9 bundle.

Previous Assumption #14:

' The plant model used in the analysis is the same as the model used for the Vermont Yankee LOCA licensig analyses.

Revised Assumotion:

The VY LdCA model was changed to accommodate the GE-13 fuel bundle. Changes in this assumption were made to adopt the revised LOCA analysis.

Justification:

The model was changed to accommodate the GE-13 fuel bundles.

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BVY 99-96 / Attachment I / Page 4

- Previous Assumotion #15:

Immediately after scram 'and isolation, the reactor pressure increase is limited by the SRVs operating in the

, pressure actuation mode. ' Without the high pressure makeup systems, the SRV actuation results in a gradual l loss of reactor water inventory. This boiloff continues with the reactor maintained at high pressure (around l 1080 psig) until the SRVs are actuated to depressurize the reactor. - The analysis included sensitivities

assuming either two orl four SRVs available for depressurization. The analysis assumed that depressurization occurs when the reacto'r water level reaches the top of active fuel in accordance with the plant emergency operating precedures.

Revised Assumption:

Immediately after scram and isolation, the reactor pres _sure increase is limited by the SRVs operating in the l " pressure actuation mode. Without the high pressure makeup systems, the SRV actuation results in a gradual l loss ofreactor. water inventory. This boiloffcontinues with the reactor maintained at high pressure (around 1113 psig) until the SRVs'are actuated to depressurize the reactor. The analysis included sensitivities assuming either two or four SRVs 'available for depressurization. The analysis assumed that

' depressurization occurs when the reactor water level reaches the top of active fuel in accordance with the

! plant emergency procedures. .

1 Justification: I l

' The transient summary is identical to the one presented in Reference'(a), with the exception of the vessel l.

pressure. The reactor vessel w~ill be maintained at the pressure corresponding to the 3% setpoint increase in the SRV with the lower setpoi.~.t (i.e.1113 psig) previously discussed in Revised Assumption #9.'

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Docket No. 50-271 BVY 09-96 1

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l Attachment 2 1

Vermont Yankee Nuclear Power Station  !

Update to Previously-Submitted RELAPS 1 Analytical Assumptions for Appendix R '

Summary of Vermont Yankee Commitments l

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SUMMARY

OF VERMONT YANKEE COMMITMENTS BVY AO.: BVY 99-96 The following table identifies commitments made in this document by Vermont Yankee. Any other actions discussed in the submittal represent intended or planned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments.

I COMMITMENT COMMITTED DATE OR " OUTAGE" None N/A i

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