B11437, Application for Amend to License DPR-65,consisting of Proposed Revs to Tech Specs for Preliminary Reload Safety Analysis Re Cycle 7 Refueling

Application for Amend to License DPR-65,consisting of Proposed Revs to Tech Specs for Preliminary Reload Safety Analysis Re Cycle 7 Refueling

ML20107A006
Person / Time
Site:	Millstone
Issue date:	02/06/1985
From:	Counsil W, Sears C NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20107A010	List: B11437, Application for Amend to License DPR-65,consisting of Proposed Revs to Tech Specs for Preliminary Reload Safety Analysis Re Cycle 7 Refueling ML20107A012
References
B11437, TAC-56814, NUDOCS 8502190222
Download: ML20107A006 (20)

Text

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n y g g General Offices e Seiden Street, Berhn, Connecticut 9 =isE Esits * "" ARTFORD, CONNECTICUT 06141-0270 k ' J ',<mema rea m. w""" "."J ",1' *C.~"". (203) 666-6911 February 6,1985 Docket No. 50-336 Bil437 Director of Nuclear Reactor Regulation Attn: Mr. James R. Miller Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) W. G. Counsil letter to R. Reid, dated March 6,1980.

(2) R. A. Clark letter to W. G. Counsil, dated June 22,1981.

(3) R. A. Clark letter to W. G. Counsil, dated January 12, 1982.

(4) R. A. Clark letter to W. G. Counsil, dated February 18, 1982.

(5) W. G. Counsil letter to R. A. Clark, dated November 17, 1983.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Cycle 7 Refueling - Preliminary Reload Safety Analysis Proposed Revisions to Technical Specifications The preliminary Reload Safety Analysis, submitted in support of the Millstone Unit No. 2 Cycle 7 reload, is attached. This report presents preliminary information concerning the Cycle 7 reload. Coolant activity measurements in Cycle 6 indicate a potential for changes to the fuel inventory expected for Cycle 7. Due to the uncertainty in the fuel inventory, the final Cycle 7 Reload Safety Analysis cannot be performed until after the shutdown of Cycle 6 when the determination of the exact fuel inventory available for use in Cycle 7 is made. The purpose of the attached report is to provide a preliminary description of the expected characteristics of the Cycle 7 reload, and to provide the bases for all changes to the Technical Specifications anticipated at this time. The Reload Safety Analysis will be submitted after the Cycle 7 reload design has been completed.

In Reference (1), Northeast Nuclear Energy Company (NNECO) provided the NRC Staff with the Basic Safety Report (BSR). The BSR serves as the reference fuel assembly and safety analysis report for the use of Westinghouse fuel at Millstone Unit No. 2. References (2), (3), and (4) document the Staff's acceptability of this report. In Reference (5), NNECO presented the Staff with 8502190222 850206 PDR ADOCK 05000336 P ppg [~[*f

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W ^..o the ' Millstone Unit No. 2 2, Cycle 6,- Reload Safety . Analyses. The BSR, as supplemented by. Reference (5), provides the basis against which the Cycle 7 reload has been preliminarily evaluated.

Cycle -7 operation will' necessitate certain changes to the Plant Technical Specifications. Therefore, pursuant to 10 CFR 50.90, NNECO hereby proposes to amend its operating license, DPR-65, by incorporating the revisions identified in the Attachment into the Millstone. Unit No. 2 Technical Specifications.1 These revisions reflect changes in Cycle 7 operating characteristics. The proposed changes -to the Millstone Unit No. 2 Technical Specifications: modify. the allowable region of operation when the core power distribution is monitored by.

the Excore Detector Monitoring System.- A new curve for the allowable thermal ~ ,

power vs axial shape index has been developed for the case when the total radial peaking factor (Fxy) is less than 1.62.. This curve allows a wider range of +

operation than the current curve developed for the case when Fxy is less than 1.719. The lower value of Fxy allows for operation at a higher thermal power and a larger axial shape index. The change establishes two curves to be used' ,

when the core power distribution is monitored 'by the Excore Detector '

Monitoring System. The new curve applies for Fxy values less than or equal to l.62 and the current curve applies -for Fxy values less than or equal to 1.719.

The parameter "N" has been removed from surveillance 9.2.1.2.c because the relationship between the allowable value of Fxy and power level is already included in the axial shape index monitoring tents.

The curve of allowable thermal power vs axial shape index is used to assure that peak linear heat rate assumed in the LOCA analysis is not exceeded.. For Millstone Unit No. 2, the maximum allowable peak linear heat rate is 15.6 kw/f t.

The proposed changes trade range in radial peaking for more range in axial shape index. The maximum radial peak is specified in the Technical Specification as a limit on Fxy. The maximum axial peak is specified by limits on the axial shape index. The allowable axial shapes will still assure that the limit on maximum peak linear heat rate of 15.6 kw/f t. is met. The increase In' the allowable value of the axial shape index will be offset by a decrease in the allowable value for Fxy without changing the design basis value for linear heat rate. Since the design basis value for linear heat rate is unchanged, safety analyses involving

!!near heat rate are not impacted by the change.

The. changes also .have no impact upon non-LOCA transients. The curve in Technical Specification 3.2.6 (Figure 3.2-4) provides the shapes to be input to all DNBR design basis analyses. Since the allowable thermal power vs axial shape index in the modified Technical' Specification is still bounded by the curve in-Technical Specification 3.2.6, the transients for which DNBR is a concern are ,

unaffected by the change. Additional justification for the changes is included in the Attachment.

The attached proposed changes have been reviewed pursuant to 10 CFR 50.59.

i and have not been found to constitute an unreviewed safety question. The

, analyses discussed herein support this conclusion in that none of the criteria of .

l 10 CFR 50.59(a)(2) are compromised. Specifically, the proposed changes do not l

, impact the previously derived maximum' allowable linear heat rate or other  ?

( parameters which could adversely impact plant transient or accident analyses.

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NNECO has keviewed the attached proposed license amendment pursuant to the requirements of '10 CFR 50.91(a) and has determined that the. changes do not involve a significant hazards determination. The basis for this conclusion is that' none of the criteria delineated in 10 CFR 50.92 have been compromised. That is,-

the proposed changes do not involve a significant increase in the probability or ' 't consequences of an accident previously evaluated; or create the possibility of a

,P new or 'different kind of accident from any accident previously evaluated; or. -,

-involve a significant reduction in a margin of- safety. A comparison of the

. contents of this amendment request with the list of examples of amendments in-

48 FR 14870 not likely-to involve significant hazards considerations reveals that

[

example (111) is applicable, in that the changes' proposed are the result of a core

. reloading and no fuel assemblies : significantly . different from those found '

previously acceptable to the NRC for previous cores at Millstone Unit No. 2 are.

3 involved. No significant changes were made to the acceptance criteria for the .

I Technical- Specifications, the analytical, methods . used to demonstrate conformance with' the , Technical Specifications Land- regulations are not

[ significantly . changed, and the . NRC has previously ' found such methods acceptable as documented in References (2) through~(4). As described above, t i,

previously approved methods have been utilized to trade margin between axial eaking~ factor to improve operational flexibility. It shape indexthat is also noted and total none radial p' examples provided as amendments likely to involv of the

! significant hazards considerations are applicable to this proposal.

I i The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the

attached proposed changes and concurs with the above determinations. -

i .

j in accordance with the requirements of 10 CFR 50.91(b), a copy of this document is being provided to the State of Connecticut.

in accordance with 10 CFR 50.170.12(c) an amendment application fee of $150 is s j enclosed with this amendment request. i-j We remain available to assist the Staff by any means to facilitate the review of

the attached proposed changes. It is requested that these changes be approved L

} prior to startup from the upcoming refueling outage, currently estimated 'to .

occur during June,1985. - We anticipate submittal of the Final Reload Safety Analysis Report on or about May 15,1985.

t Very truly yours, 4

I NORTHEAST NUCl. EAR ENERGY COMPANY e

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W. G. Counsil Senior Vice President ,

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! Vice President -

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' CCt Director, Radiation Control Unit Department of Environmental Protection .,

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- State Office Building .

Hartford, CT 06116 STATE OF CONNECTICUT - )

) ss. Berlin COUNTY OF HARTFORD )

1 Then personally appeared before me C. F. Sears, who being duly sworn, did state that he is 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 are true and correct to the best of his knowledge and belief, be e ll i i '

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O Docket No. 50-336 Attachment Millstone Nuclear Power Station, Unit No. 2 Cycle 7 o Cycle 7 Preliminary Reload Safety Analysis o Justification of Technical Specification Revisions o Proposed Revisions to Technical Specifications February,1985

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l TABLE OF CONTENTS Section Title 1.0 Introduction 1.1 ' Objectives 1.2 Fuel Inventory for Cycle 7 2.0 ' Cycle 7 Preliminary Physics Characteristics '

2.1 Normal Inventory Reload Design

'2.2 Potential Impact of Inventory Changes 3.0 Justification for Technical Specification Changes 3.1 Hardware Change to Lead CEA Bank 3.2 Total Planar Radial Peaking Factor and Linear Heat Rate Monitoring 4.0 References Appendix Technical Specification Changes 1

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LIST OF TABLES ,

Table . Title

-1 Shutdown Requirements and Margins 2 Rod Ejection Accident Analysis Results ,.

3 Sequence of Events - CEA EJeetion incident s

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LIST OF FIGURES Figure Title 1 Millstone Safety Analysis Rod Ejection Incident - HFP/ Nuclear Power Versus Time 2 Millstone Safety Analysis Rod Ejection Incident - HFP/ Fuel Center Temperature Versus Time 3 Millstone Safety Analysis Rod Ejection incident - HFP/ Fuel Average Temperature Versus Time 4 Millstone Safety Analysis Rod Ejection incident - HFP/ Clad Temperature Versus Time 1

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t 1.0 Introduction . ..

' 1.1= Objectives .

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' This report presents preliminary information concerning the core reload of L the Millstone Nuclear Power Station Unit No. 2, Cycle 7. : Coolant _ activity -

measurements .in . Cycle 6 Indicate a potential for changes to the fuel inventory expected 'for : Cycle '7. _ Due- to the ' mcertainty in the fuel inventory, the Cycle 7 reload safety analysis .will be performed af ter_ the '

shutdown of Cycle 6 and the determination of the exact ' fuel inventory.

available for use in Cycle 7 Is made. The purpose of this report is to provide a preliminary description of the expected characteristics of the' i Cycle 7 reload, and to provide the bases for all changes to the Technical  !

Specifications anticipated at this tin.e. The Reload Safety Analysis Report will be submitted af ter the Cycle 7 reload design has been completed.

1.2 Fuel Inventory for Cycle 7-The feed fuel for the Millstone Unit No. 2, Cycle 7 core will consist of 24 split-enrichment interior feed assemblies, each containing 60 fuel rods at l 2.6 w/o and 116 fuel rods at 2.9 w/o, and 48 split-enrichment peripheral feed assemblies, each containing 60 fuel rods at 2.9 w/o and 116 fuel rods at 3.3 w/o. The feed fuel will replace 20 Combustion Engineering (CE)

Batch A assemblies,1 CE Batch B assembly, and 51 Westinghouse Batch F assemblies which will be discharged from the core at the end of Cycle 6.

An additional 5 Westinghouse Batch F assemblies will be dischargrd at the end of Cycle 6 and will be replaced by 5 Batch F assemblies which were ,

removed from the core at the end of Cycle 5. Twenty-four CE Batch A assemblies from Cycle I will .also be available for use in Cycle 7.

Additional fuel inventory will be available from the reconstitution of fuel assemblies which were removed from the core at the end of Cycle 5 with -

indication of f ailed fuel rods (Ref.1).

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a 2.0 Cycle 7 Preliminary Physics Characteristics

'2.1 Normal Inventory Reload Design A core loading pattern was developed for the Cycle 7 reload based on the

-assumption of no fuel failure in Cycle 6. The parameters which have been historically the most limiting for Millstone Unit No. 2 were analyzed using this loading pattern,' namely, the Moderator Temperature ' Coefficient (MTC), radial peaking factor, and the available Shutdown Margin (SDM) at the most ilmiting condition during the cycle. The current limits on MTC,-

radial peaking factor, and SDM were met with this reload design. Table 1 provides the control. rod worths and requirements at the most limiting condition during the cycle. -

2.2 Potential Impact of Inventory Changes In order to provide early identification of potential problem areas in the Cycle 7 reload design, loading pattern scoping studies were performed based on three different assumed fuel inventory scenarios. The basic assumption of the scoping studies was that the extent of fuel damage in Cycle 6 is similsr to that of Cycle 5, as indicated by the coolant activity measurements. The three scenarios assumed different distributions of the -

fuel failure in Cycle 6'in order to simulate the effects of a variety of possible reactivity and burnup distributions. Loading patterns were established for each of the three scenarios, and the MTC, radial peaking factor (Fr), total planar radial peaking . factor (Fxy), and SDM were analyzed. In comparison with past behavior of Millstone Unit No. 2, the parameter which was most affected by the atypical fuel inventories of the scoping studies was the planar radial peaking factor. The best estimate Fxy calculated in the scoping studies showed an increase of approximately .

3 percent over the values determined in previous cycles. . Based on the results of these three studies, a Technical Specification change on Fxy is not anticipated; however, the Cycle 7 loading pattern based on the actual fuel inventory may necessitate changes.

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3.0 Justification for Technical Specification Changes Analyses were performed in order to justify Technical Specification changes proposed at the present time for Millstone Unit No. 2, Cycle 7. The results of these analyses are presented here.

3.1 Hardware Change to Lead CEA Bank In anticipation of hardware changes to be made to the lead Control Element Assembly (CEA) bank during the plant outage following the Cycle 6 shutdown, an evaluation was performed to determine the impact of the hardware changes on the current Technical Specifications and safety analysis inputs. The part-strength control rods in the lead CEA bank are to be replaced by full-strength control rods, making the lead bank CEAs identical in composition to the remainder of the CEA banks.

Parameters which are input to the safety analysis and which would be affected by the proposed hardware change were analyzed. These parameters are radial peaking factor, dropped rod peaking factor, ejected rod worth and peaking factor, and shutdown margin. The only input parameter to the safety analysis which exceeded the current limit established by the Basic Safety Report (BSR) (Ref. 2) and the Cycle 6 reload safety analysis (Ref. 3) was the HFP ejected rod worth.

The ejected rod accident at HFP was reanalyzed using the methodology described in Reference 4. The parameters used in the analysis are given in Reference 5, with the exception of the ejected rod worth and the reactor coolant flow. The reactor coolant flow used is given in the Cycle 6 reload safety analysis (Ref. 3). The value of the ejected rod worth used in this analysis was 0.28% . The results of this analysis are given in Table 2.

The sequence of events for this accident is given in Table 3. The nuclear power transient and the hot spot fuel and clad temperature transients are shown in Figures 1 through 4. The results demonstrate that the limiting criteria given in Reference 2 for this accident are not exceeded. The average enthalpy of the hottest fuel pellet does not exceed the damage threshold of 200 cal /gm.

An analysis was performed in order to verify the applicabl!!ty of the current Technical Specifications on the Axial Shape index alarm setpoints for operation using the excore detector monitoring system, the fuel center-line melt trip, the Limiting Condition for Operation (LCO) on ASI for Departure from Nucleate Bolling (DNB), and the Thermal Margin / Low Pressure (TM/LP) trip. The Condition I and 11 power shapes were analyzed using the methodology described in Reference 4 with the proposed rod configuration. The results of these analyses showed that the proposed hardware change does not require changes to these Technical Specifications.

3.2 Total Planar Radial Peaking Factor and Linear Heat Rate Monitoring The Axial Shape index (ASI) envelope currently used to monitor the linear heat rate while operating on excore detectors was generated using the power-dependent radial peaking factor (Fxy) given in Technical

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. Specification Figure 3.2-3 in accordance with the methodology described in Reference 2. The maximum Fxy . allowed by the . current Technical Specification is' 1.719 at full power.-- The current en'velope was verified to be unaffected by the proposed hardware change to the lead CEA' bank as

' described in Section 3.1. An additional analysis'.was performed using the power-dependent Fxy shown in Figure 5, with a maximum of 1.62 at full *

. power.- The lead CEA bank hardware change was accounted for in the additional analysis. Given the Fxy relationship of- Figure 5, the power shape analysis showed that it is acceptable to expand the ASI operating

' envelope to the limits shown in Figure 6. . It :Is - proposed that the surveillance requirements .on the -linear heat rate and the- Technical-Specification on the Total Planar Radial Peaking Factor (Fxy) be modified

. as given in the ' Appendix to allow the use of the wider operating range when Fxy is 61.62.

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i 4.0 References ~i

1. Osborne, D., " Meeting Summary", Notes of 'NUSCO/NRC Meeting' of October 3,1984 at Bethesda, MD, October 26,1984.

2. . Millstone Unit No. 2, " Millstone Unit No. 2 Basic' Safety Report", Docket No. 50-336, March,1980.

3. W. G. Counsil letter to 3. R. Miller, dated November 4,1983..

4. Bordelon, F. M ., et. al., . " Westinghouse Reload Safety Evaluation Methodology", WCAP-9273, March,1978.

5. W. G. Counsit letter to R. A. Clark, dated November 17,1981.

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SHUTDOWN REQUIREMENTS AND MARGINS i MILLSTONE UNIT'2 - CYCLE 7-

- NORMAL FUEL INVENTORY ASSUMED Control Rod Worth (%Ao)- EOC 7 All Rods Inserted 8.36 All Rods Inserted Less Worst Stuck Rod 6.86 (1) Less 10% 6.17 Control Rod Requirements Reactivity Defects '(combined Doppler, Tavg, Void, and Redistribution Effects) 2.58 Rod Insertion Allowance 0.42 (2) Total Requirements 3.00 '

ShutdownMargin((!)-(1))(%ao) 3,17 Required Shutdown Margin (%ao) 2.90 i a

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TABLE 2 RESULTS OF THE CEA EJECTION ACCIDENT ANAL) SIS i

HFP Max. fuel pellet average temperature. 'F 4228 t

Max. fuel center temperature. *F 5011 Max, clad average temperature. *F 2465 Max. fuel pellet center melting, percent 8.44 Max fuel stored energy, cal /gm 186

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0.04 High Power Trip Signal Generated 112 percent 0.1 CEA Fully $jected -

0.13 Peak' Nuclear Flux Reached See Fig. 1 0.94 CEA Insertion Begins -

0.36 Peak Fuel Temperature Reached See Fig. 2 3.54 CEA's Reach 90 percent Insertion -

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