ML20128B281

Proposed Tech Specs Consisting of Proposed Change Request 92-19,revising TSs MCPR Safety Limits Since Core Will Be Reloaded W/New Fuel Type,Ge 11 Fuel

ML20128B281
Person / Time
Site:	Peach Bottom
Issue date:	01/28/1993
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20128B278	List: ML20128B277 ML20128B281
References
NUDOCS 9302020474
Download: ML20128B281 (7)

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l ATTACHMENT ..o PEACH BOTTOM ATOMIC POWER STATION UNIT 3  !

I Docket No. 50-278 License No. DPR-56 TECHNICAL SPECIFICATIONS CHANGE REQUEST.

No. 92-19

" Minimum Critical Power Ratio Safety Limits" 9302020474 930128 PDR ADOCK 05000278

.P. .PDR

i Docket No. 278 License No. DPR-56 INTRODUCTION Cycle 10 operation of Peach Bottom Atomic Power Station (PBAPS Unit 3 necessitates revision of the Technical Specifications (TS)

Minimum Critical Power Ratio (MCPR) Safety Limits since the core will be rsloaded with a new fuel type, GE 11 fuel. Unit 3 Cycle 10 is scheduled to begin in November of 1993.

Philadelphia Electric Company (PECo) hereby requests that, once approved, these changes be effective August 16, 1993 for Unit

3. This date is required to accommodated the use of GE 11 fuel in PBAPS Unit 3 and to support the associated reload licensing calculations.

DESCRIPTION OF CHANGES:

The Licensee proposes that the Safety Limit, Section 1.1.A;

" Reactor Pressure 2 800 psia and Core Flow 2 10% of Rated," be revised to reflect the new limits for Gell fuel. The present wording would remain the same except 1.07 and 1.08 would replace the present values of 1.06 and 1.07 for two loop and single loop operation, respectively. The proposed words would read: "The existence of a minimum critical power ratio (MCPR) less than 1.07 for two loop operation, or 1.08 for single loop operation, shall constitute violation of the fuel cladding integrity safety limit."

SAFETY DISCUSSION The current Unit 3 TS MCPR Safety Limits are 1.06 for two-recirculation loop operation and 1.07 for single recirculation loop operation (page 9 of TS). However, use cf GE 11 fuel in Unit 3-during Cycle 10 requires MCPR Safety Limits not less than 1.07 for two-loop operation and 1.08 for single loop operation.

The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal hydraulic conditions resulting in a departura from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rod, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of critical power. Therefore, the fuel cladding integrity safety

Docket No.-278 License No. DPR-56 limit is definea as the critical-power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core _ are expected- to avoid boiling transition considering-- the power distribution within the core and all uncertainties, u

The SLMCPR is determined using the NRC approved ~ General '

Electric Thermal - BasA s ( GEi'AB ) described in' " General Electric Standard application for Reactor Fuel," NEDE-240ll-P-A-10, February, 1991 and " General Electric BWR Thermal Analysis (GETAB):

Data Correlation and Design Application," NEDO-10958-A, January 1977, for two recirculation loop operation.. The SLMCPR is increased by 0.01 for single loop operation as described in

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

The SLMCPR is influenced by the critical power correlation and by bundle design parameters which affect the- bundle R-factor d'.stribution and the core radial power distribution.- These parameters include the spacer design assembly dimensional geometry, unrichment level and distribution, and fuel discharge exposure.

For the Gell fuel design, there are significant design changes from previous designs, thereby requiring a recalculation of the SLMCPR.

A Safety Limit MCPR of 1.07 (1.08 for single loop) has been approved by the NRC for D- or C-lattice plants. operating with a reload core of Gell fuel (Gell Compliance with Amendment 22 of NEDE-240110-P-A GESTAR 11, NEDE-31917P, April 1991). PBAPS Unit 3 is a D-lattice plant and the reload fuel for Cycle 10 is of the Gell design. Approximately, one-third of the core will be replaced with fresh GE 11 bundles. The only exception to the PBAPS Unit 3 Reload 9 (Cycle 10) batch is that four bundles of the SPC 9x9A design will'be included. The-four Lead Use Assemblies (LUAs) will-be loaded in non-limiting locations such that the LUAs will have an insignificant impact on the core wide SLMCPR. The LUAs will be:

evaluated for applicability of a SLMPCPR of 1.07.

As discussed ~previously, the proposed MCPR Safety Limits have been established in accordance with NRC-approved methods. In.

addition, . conservative MCPR operating limits will also be established using NRC-approved methods in accordance with ,TS 6.9.1.e(l) and (2) and will be published in the Core Operating i Limits Report (COLR) for Cycle 10. The COLR will be submitted to the NRC upon issuance in accordance with TS 6.9.1.e(4).

Docket No. 278 License No. DPR-56 The accidents previously evaluated which are potentially impacted by this change are the limiting Anticipated Operational Occurrences ( AOos) specifically analyzed for each operating cycle.

These AOOs are Rod Withdrawal Error, Loss of 100*F Feedwater Heating, Generator Load Rejection Without Bypass, Feedwater Controller Failure, Fuel Loading Error, and Rotated Bundle Error.

These events are described in the United States supplement to GESTAR.

PECo proposes that the changes to the MCPR Safety Limits do not involve significant hazards considerations for the following reasons.

1) The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Because the MCPR Safety Limits are operational thresholds analytically selected using proven methods, they cannot, themselves, initiate an accident. The probacility of occurrence of transients is determined by the f requency of operator errors and equipment failures, not by the adequacy of the MCPR Safety Limits selected. Because the proposed MCPR safety limits have been selected such that no fuel damage is calculated to occur during the most severe moderate frequency transient events, they will ensure that the consequences of these events are not increased. The response of the plant to transients will be within the bounds of the discussion in Chapter 14 and Appendix G of the Updated Final Safety Analysis Report since the proposed MCPR Safety Limits will accomplish the same objectives as the previous limits.

11) The proposed changes do-not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed MCPR Safety Limits have been selected such that the design basis _is satisfied. The MCPR Safety Limits are operational thresholds analytically selected using proven-methods; therefore, they cannot, themselves, initiate an accident.

An improperly selected limit could result in fuel damage, which is a consequence of previously evaluated accidents.

Thus, no new or different type of accident could be created by revising the limits.

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Dock'et No'. 2782

-License No. DPR-564

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lii) The proposed- changes do inot- involve a significanti l

'l reduction in a margin of safety because_the proposed MCPR; ,

Saf ety- . Limits . have. been- selected i such that'- the _ design- y!

basis = is satisfied and- such that; the:-conservatism-described-in'the' Bases for the Fuel Cladding Integrity- I Safety-Limit TS are-maintained. Thus,= margins ofisafety-with the proposed MCPR Safety-Limits: are the same as.with the previous limits.

ENVIRONMENTAL IMPACT l 1

An ~ environmental assessment: is not; required'for the changes--  :

requested by this Application because the requested changes .conformL 4 to the criteria for _" actions eligible for categorical exclusion" as ' .

specified -in 10CFR51.22 (c)(9 ) .- The _-_ requested ' changes ~ hav.e _ bee _n shown by this Application not-to adversely affect the systems and equipment that prevent ~ the _ uncontrolled release - 'of : radioactive - -

material to the- environment. The -Application _ involves -

no

- significant hazards considerations as demonstrated in the preceding.

sections. The Application involves noTsignificant-change 11n the-types or significant. Increase _in the amounts of any ef fluents: that-' a =

may be released offsite, and'there-will:be no significant increase in individual or cumulative occupational radiationrexposure. .

a CONCLUSION-a The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed - changes to= the.. Technical Specifications- and determined that they 'do_ _not , involve. an Unreviewed-Safety Question-and-will;not endanger-'the health and' safety of the public.

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ATTACHMENT 2 PEACH BOTTOM ATOMIC POWER STATION UNIT 3 Docket No. 5-278 License No. DPR-56 REVISED TECHNICAL SPECIFICATION PAGES List of Attached Pages Unit 3 9

Unit 3 i PBAPS SAEETY LI141T .

LIMITING SAFETY _jlySTEM SETTING 1.2 EVEh CLADDlHG_IRIEGRITY 2.1 EUEL CLADDING-INTEGRITY SDDLLCabilitX: hpplicitbi1ity:

The Safety Limits established The Limiting Safety System to preserve the fuel cladding Settings apply to trip integrity apply to those settings of the instruments variables which monitor the and devices which are provided fuel thermal behavior. to provent the fuel cladding integrity. Safety Limits from Rhientiynas being exceeded.

The objective of the Safety Qbjectivest Limits is to establish limite which assure the integrity of The objective of the Limiting the f.uel cladding. Safety System Settings is to define the level of the proceso variables at which Sp_qc111n tinnt automatic protective action is initiated to prevent the fuel A. Reactnr fJ.cssure d BSD psia cladding intagrity Safety auifore Flow a 1AL_o1 Limits from being exceeded.

Ritted Specification:

The existence of a minimum critical power ratio (MCPR) The limiting safety system l less than 1.07 for two settings shall be as specified recirculation loop below:

l operation, or 1.00 for single loop operation, A. Neutron Flux Scram shall constitute violation of the fuel cladding 1. APRM Plux Ser m Trip integrity safety limit. Setting _(Run Mode)

To ensure that this safety When the Mode Switch is in limit is not exceeded, the RUN position, the APRM neutron flux ehall not be flux scram trip setting above the scram setting shall bei established in specification 2.1.A for Ss 0.58W + 62% - 0.58tiW longer than 1.15 seconds as indicated by the process where:

computer. When the process computer is out of service S- Setting in percent of this safety limit shall be rated thermal power assumed to be exceeded if (3293 MWt) the neutron flux exceeds its scram setting and a W= Loop recirculating control rod scram does not flow rate in percent eccur. of design. W is 100 for core flow of 102.5 million lb/hr or greater.

Arendment No. 14, 41, 77, 79 150, 159

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