ML20113H191

Safety Evaluation for Tech Spec Changes Re Reload 6/Cycle 7

ML20113H191
Person / Time
Site:	FitzPatrick
Issue date:	01/16/1985
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20113H173	List: ML20113H172 ML20113H179 ML20113H186 ML20113H191 ML20113H205 ML20113H214
References
JPTS-85-002, JPTS-85-2, NUDOCS 8501250028
Download: ML20113H191 (6)

Text

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l ATTACHMENT II TO JPN-85-03 Safety Evaluation for Technical Specification Chances Related to Reload 6/ Cycle 7 (JPTS-85-002) i l

l New York Power Authority-James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 8501250028 850116 PDR ADOCK 05000333 P PDR

I. Description of the Proposed Chances On page vii (List of Figures), the list has been changed to reflect the deletion of Figures 3.5-6, 3.5-7 and 3.5-8. Figure 3.5-11 has been added to reflect the new fuel type to be loaded in Reload 6.

On page-31, Section 3.1, the table entitled "MCPR Operating Limit for Incremental Cycle Core Average Exposure" has been revised to reflect transient analyses performed for the Reload 6/ Cycle 7 core. These analyses are reported in Reference 3 (Attachment III).

Figure 3.'l-2 (" Operating Limit MCPR Versus tau (Defined in Section 3.1.B.'2) for all Fuel Types") on page 47b-has been revised to reflect transient analyses performed for the Reload 6/ Cycle 7 core. These analyses are reported in Reference 3.

Section 3.5.H (" Average Planar Linear Heat Generation Rate

'(APLHGR)") on page 123 has been revised to eliminate references to figures deleted by other portions of this proposed change.

On page 130, Section 3.5.H (" Average Planar Linear Heat Generation Rate (APLHGR)") is revised-to eliminate references to figures which are deleted by this application.

Figures 3.5-6, 3.5-7 and 3.5-8 (on pages 135d, 135e and 135f respectively) have been replaced with-blank figures. These three figures were " Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Planar Average Exposure" for three different fuel types (Reload 2-8DRB283, Reload 3-P8DRB265L and Reload 3-P8DRB283).

The-three fuel types associated with these three figures will be discharged as part of the forthcoming Reload 6/ Cycle 7 refueling.

Therefore,- these figures are no longer necessary.

Figure 3.5-11, " Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Planar Average Exposure, page 1351) has been added.- This new figure represents the new fuel'to be loaded-in Reload 6 (See Reference 7).

II. Purpose of the Proposed Chances The changes made to page vil are purely administrative; they update the List of Figures to-reflect changes made as part of this proposed change.

The changes to the Section 3.1 table on page 31 ("MCPR Operating Limit for Incremental Cycle = Core Average Exposure")

reflect cycle specific transient analyses performed by General Electric for-the Reload 6/ Cycle 7 core (See References 3, 4 and 7).

The new Figure 3.5-11 represents MAPLHGR limits for the new fuel type to be added during Reload 6. Figures 3.5-9 and 3.5-10 apply to the fuel types that will remain in the core from previous cycles.

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Although fuel type BP8DRB299 is identical to Reload 5 fuel in nuclear design, Reload 6 fuel has been fitted with eighty mil fuel channels rather than the one hundred mil channels used for previous reloads. This change in channel thickness results in a slightly different fuel bundle response during a loss-of-coolant accident ,

(LOCA) in the high exposure range. Consequently, different MAPLHGR limits are applied to Reload 6 fuel.

III. Impact of Proposed Chances These proposed changes are necessary for the forthcoming Reload 6/ Cycle 7 reactor refueling currently scheduled to. start February 16, 1985. The Technical Specification changes proposed by this' application are necessary to account for new fuel to be added to the reactor core as well as those fuel types to be discharged and the effects of these changes on plant analyses.

Reference 3 summarizes the analyses performed by General Electric in support of the changes proposed by this application.

This report supplements General Electric's generic report, Reference

4. The generic report (GESTAR) details the safety analyses for mechanical and nuclear fuel design, as well as methodology for transient and accident analyses. The supplemental report (Reference

3) documents the results of General Electric's analysis for the FitzPatrick Reload 6/ Cycle 7 core. Reference 3 has been prepared using the format described in Appendix A of Reference 4. The numbers appearing in parenthesis following each section heading refer to GESTAR Section numbers where the analyses are described in more detail.

.The Commission has provided guidance concerning the application of the standards for making a "no significant hazard considerations" determination by providing certain examples in the Federal Register (F.R.) Vol.~48, No. 67 dated April 6, 1984, page 14870 (Reference 5). The proposed changes match Commission example (iii), which states in part "for a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different-from those found previously acceptable...". A single new type of fuel (BPDRB299) is planned for.this reload. This new fuel type differs from the fuel types currently in use at FitzPatrick in two: aspects; (1) it is a Barrier fuel, and (2) it is fitted with eighty mil thick fuel channels rather than the previously used one hundred mil channels. The Barrier fuel design has a zirconium layer metallurgically bonded to the inside surface of the Zircalloy-2 fuel cladding. This feature is expected to reduce the probability of pellet-clad interaction fuel failures. The Barrier fuel design has been incorporated into the current revision of GESTAR (Reference 4) and has been determined by the NRC to be acceptable for referencing i in license applications. (Reference 8). The change from one hundred ,

mil to eighty mil channels is actually a return to initial core '

' channel dimensions although subsequent reloads used one hundred mil channels.

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Since eighty mil channels have been used successfully at l FitzPatrick, and extensively on other plants similar in core and I fuel design to FitzPatrick, this is not a significant change. l The analytical methods used to demonstrate conformance with the Technical Specifications and regulations are described in Reference  ;

4. These methods have not changed significantly from the methods used for previous reload submittals. Reference 4 has been reviewed

and approved by tho NRC (Reference 8).

Operation of the FitzPatrick plant in accordance with the proposed amendments, therefore, would not:

4 (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

j IV. Implementation of the Chance The proposed changes will not impact the Fire Protection or ALARA programs at FitzPatrick, nor will they have any significant impact on the environment.

V. Conclusion The incorporation of this change: a) will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety 4 Analysia Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluted previously in the Safety Analysis Report; c) will not reduce the margin of safety as defined in the Bases for any Technical Specification: d) does not constitute an unreviewed safety question; and e) involves no significant hazards conaideration, as defined in 10 CRP 50.92.

VI. References

1. James A. FitzPatrick Nuclear Power Plant Safetf Evaluation Report (SER),

2. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev. 2, July, 1984.

3. General Electric Report, " Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant - Reload 6," 23A1806, November 1984. (Included as Attachment III).

4. General Electric Report, " General Electric Standard Application for Reacter Fuel", (GESTAR), NEDE-240ll-P-A-6, April 1983.

5. Federal Register, Vol.48 No. 67, April 6, 1983, Interim Final Rule " Standards for Determining Whether License Amendments Involve No Significant Hazards Considerations" pages 14864-14880.

6. 10 CFR 50.59, " Changes, tests and experiments."

7. General Electric Report, " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)", NEDO-21662-2, July 1977, as amended. (Included as Attachment IV).

8. NRC letter dated April 13, 1983, C.O. Thomas (NRC) to J.S.

Ci ~~91y (GE) regarding Acceptance for Referencing of Licensing Tog 1 Report NEDE-240ll-P-A-4, " Barrier Fuel Amendment to General Electric Standard Application for Reactor Fuel."

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ATTACHMENT III TO JPN-85-03 n

Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant Reload 6 (JPTS-85-002)

New York Power ~ Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333