ML20024C081

Application for Amend to License NPF-3,revising Tech Specs Re Cycle 4 Reload

ML20024C081
Person / Time
Site:	Davis Besse
Issue date:	07/05/1983
From:	Murray T TOLEDO EDISON CO.
To:
Shared Package
ML20024C071	List: ML20024C070 ML20024C081 ML20024C092
References
NUDOCS 8307120237
Download: ML20024C081 (6)

Text

I APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Facility Operating License No. NPF-3, together with the. Safety Evaluation for the requested change.

The proposed changes include Cycle 4 Reload Report (BAW-1783 May 1983).

By

/s/ T. D. Murray Station Superintendent For R. P. Crouse Vice President, Nuclear I

(-

Sworn and subscribed befori me this 5th day of July, 1983.

I l

Nora Lynn Flood i

Notary Public, State of Ohio My Commission expires Sept. 1, 1987 l

l SEAL l

8307120237 830705 PDR ADOCK 05000346 P

PDR i

I Docket No. 50-346 License No. NPF-3 Serial No. 965 July 5, 1983 Attachment i

I.

Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications A.

Time required to Implement. This change is to be effective upon NRC approval.

B.

Reason for Change (Facility Change Request 82-106 Supplement-1).

To reflect changes in Cycle 4 core.

C.

Safety Evaluation See Attached D.

Significant Hazards Considerations See Attached

SAFETY EVALUATION

'i This amendment request proposes reloading of new fuel and the shuffling l

of fuel and control rod assemblies to facilitate nuclear power generation l

for Cycle 4-in accordance with the limits and analysis presented _in the

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attached' Reload Report BAW-1783. The safety function of the Reload j

Report and the affected Technical Specifications is to ensure operation j

of the core within safety-limits.

The' reference cycle for the nuclear and thermal-hydraulic design of Cycle 4 is the projected Cycle 3 End of Cycle (EOC) conditions. The Cycle 4-1 physics' parameters are based on a 268 effective. full power day (EFPD)

Cycle 3 length including APSR withdrawal and coastdown. There were no y

- anomalies during Cycle 3 which would adversely affect. fuel performance

' during Cycle 4, as designed.1.The loading includes 48 new fuel assemblies and the reinsertion of 37'previously discharged fuel assemblies with no mechanical design changes for the reload.

The.. Cycle 4 design is characterized by only eight fuel assemblies being cross core shuffled so

- as to minimize any_ carryover effects from tilts encountered in previous

_ cycles.

4

- The Cycle 4 thermal analyses represents a change in the analytical

. method. The analyses for the incoming batch 6 fuel have been performed

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using the TACO 2 computer code. This contains the NRC imposed densifications model and results in a 0.5 kw/ft reduction from the 16.5 kw/ft LOCA limit at_the 2 ft. elevation reported in BAW-10105 due to the change in heat transfer thru the altered gap size.

It is projected that i.

there are no changes to the kw/ft limit at other elevations due to-TACO 2.

The impact of NUREG-0630 has been incorporated into the ECCS model for Cycle 4.

It contains revised models for cladding rupture, strain and blockage during and following a LOCA and results.in an c

additional 0.5 kw/ft reduction. The overall effect, then is a new LOCA.

limit of 15.5'kw/ft. -The 15.5 kw/ft analysis was only performed for-the 2 ft. elevation. This restriction applies only to the first 24 EFPD, at which time the temperature'is sufficiently reduced to allow the limits 4'

reported in BAW-10105. The new limits translate to more stringent control rod insertion and imbalance limits for'the first 24' effective full power days (EFPF) of operation..

The thermal-hydraulic design results in Cycle 4 do not differ

. significantly from that in Cycle 3.

The moderator and Doppler coefficients remain negative for Cycle 4.

The moderator coefficient is less negative than the FSAR value so that the FSAR value is bounding for main steam line break or any overcooling transients. The maximum dropped rod worth is less than the FSAR value which compensates for the slightly more negative Doppler coefficient at BOC4. The ejected rod worth is'also smaller than the FSAR value. The boron reactivity worth is also bounded by the FSAR value for the moderator dilution accident. The FSAR accidents have been examined by B&W with respect to the Cycle 4 4

- parameters'to ensure that the thermal performance during the hypothetical 4

transients has not been degraded, n

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SAFETY EVALUATION a

This amendment request proposes reloading of new fuel and the shuffling of. fuel and control rod assemblies to facilitate nuclear power gene. ration

for Cycle 4 in accordance with the limits and analysis presented in the attached Reload Report'BAW-1783. The safety function of the Reload

.Renort and the affected Technical Specifications is to ensure operation of the core within safety limits.

y L

The reference cycle for the nuclear and thermal-hydraulic design of Cycle 4 is the projected-Cycle 3 End of Cycle (EOC) conditions. The Cycle 4 physics parameters are based on a 268 effective full power day (EFPD)

Cycle 3 length including APSR withdrawal and coastdown. There were no

~

anomalies during Cycle.3.which would adversely affect fuel performance during Cycle 4, as designed. The londing includes 48 new fuel assemblies and the reinsertion of 37 previously discharged fuel assemblies with no mechanical-design changes for the reload. The Cycle 4 design is characterized by.only eight fuel assemblies being cross core shuffled so as to minimize any carryover effects from tilts encountered in previous

[

cycles.

. Thi Cycle 4 thermal analyses represents a change in the analytical'

. method. The analyses for the incoming batch 6 fuel have been performed.

using the' TACO 2 computer code. This contains the NRC imposed j

densifications model and results in a 0.5 kw/ft reduction from the 16.5 kw/ft LOCA limit at the 2 ft. elevation reported in BAW-10105 due to the change in heat transfer thru the altered gap size.

It is projected that there are no changes to the.kw/ft limit at other elevations due to

. TACO 2.

The impact of NUREG-0630 has been incorporated into the ECCS c

model for Cycle 4.

It conta s revised models for cladding rupture, strain and blockage during t;he, following a LOCA and results -in an additional 0.5 kw/ft reduction. The overall effect, then is a new LOCA limit'of 15.5 kw/ft. The 15_.5 kw/ft analysis was only performed for the 2-ft.. elevation.- This restriction applies only to the first 24 EFPD, at which time the tewperature is sufficiently reduced to allow the limits reported in BAW-10105. The new limits translate to more stringent control rod insertion and imbalance limits for the first 24 effective full power days (EFPF) of operation.

The thermal-hydraulic design results in Cycle.4 do not differ significantly from that'in Cycle 3.

The moderator and Doppler

' coefficients remain negative for Cycle 4.

The moderator coefficient is less negative than the FSAR value so that the FSAR value is bounding for

~

main steam line break or any overcooling transients. The maximum dropped rod worth is less than'the~FSAR value which compensates for the slightly more negative Doppler coefficient at BOC4..The ejected rod worth is also smaller,than the FSAR value. The Foron reactivity worth is also bounded by the FSAR value for the moderatr.r dilution accident. The FSAR accidents have been examined by T&W with respect to the Cycle 4 parameters to ensure that the charmal performance during the hypothetical

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transients has not been degradet.

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. SAFETY EVALUATION (CON'T)

The pertinent Technical Specifications in the Reload Report have been revised for Cycle 4 operation to account for changes in power peaking and control rod worths. The planned start-up physics test program is the same as that' performed for Cycle 3 and is sufficient to demonstrate that the core will perform within the assumption of the safety analysis. The core power distribution tests will be used as a check on the power distribution in the core to identify any abnormalities and to maintain the quadrant power tilt within the Technical Specifications limits.

Therefore, it is concluded that the changes proposed in the Cycle 4 Reload' Report do not involve an unreviewed safety question.

i i

SIGNIFICANT HAZARD CONSIDERATION The proposed license amendment request, Cycle 4 Reload Report, for the Davis-Besse Nuclear Power Station Unit No. I does not contain a significant hazard. The Reload Report is dimilar to previously submitted reload reports in design and limits considerations. The previous reports have been evaluated and did not invoke a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, this request meets 10 CFR 50.92 -(c)(1).

The analysis of Cycle 4 Reload Report did not identify any new or

..The different kind of accident from any previously evaluated by the NRC.

accident and consequences are bounded by previously evaluated design and submitted Reload Report. Therefore, this amendment request does not create the possibility _ of a new or different accident from any accident previously evaluated and meets 10 CFR 50.92 (c)(2).

The evaluations of the reload report that all margins of safety are bounded by previous evaluations and analysis. Any change to setpoint or margins are'small and within licensed design limits for Davis-Besse.

Therefore, we meet Section 50.92 (c)(3).

In accordance with the issued rule, the Commission notes examples of Amendments that are considered not likely to involve Significant Hazard Considerations which includes changes resulting from Reload Reports.

There are no new fuel assemblies significantly different from previously accepted by the NRC. There are no significant changes made to the acceptance criteria and analytical methods used to demonstrate l

conformance with Technical Specifications. The rules and regulations have not been significantly changed and the NRC has previously found such methods acceptable.

Therefore, based on the attached safety evaluation and the above, the requested amendment'does not contain a Significant Hazard, f

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BAW-1783 May 1983 l

l DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 4 - RELOAD REPORT l

1 Babcock &Wilcox

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