ML20005G276

Proposed Tech Specs Re Safety Limit Min Critical Power Ratio

ML20005G276
Person / Time
Site:	FitzPatrick
Issue date:	01/09/1990
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20005G268	List: ML20005G267 ML20005G274 ML20005G276
References
JPTS-89-036, JPTS-89-36, NUDOCS 9001180377
Download: ML20005G276 (6)

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ATTACHMENT l

i o PROPOSED TECHNICAL SPECIFIC TION CHANGES REGARDING -

SAFti V UMIT MINIMUM CRITICAL POWER RATIO.

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JAMES A. FITZPATRICK NvCLEAR POWER PLANT Docket No. 50-333 l

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1.1 FUEL Cl. ADDING INTEGRITY 2.1 FUEL Ct. ADDING INTEGRl'iY '

Applicability: Applicability-The Safety Umits established to preserve the fuel cladding integrity The Umiting Safety System Settings apply to trip settings of the instruments and devices which are provided to prevent the fuel cladding apply to those variables which monitor the fuel thermal behavior.

Objective:

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ective:

The objective of the Safety Umits is to establish limits below which the The objective of the Umiting Safety System Settings '.s to define the level integrity of the fuel cladding is preserved.

of the proce;,s variables at which automatic protective action is irwtiated to prevent the fuel cladding integrity Safety Umits from being exceeded.

Specifications:

Spacifications:

A. Trip Settings A. Reactor Pressure >785 psig and Core Flow > 10% of Rated The limiting safety system trip settings shall be as specEisd The existence of a minimum critical power ratio (MCPR) less below-than 1.07 shall constitute violation of the fuel cladding integrity l l 1. Neutron Flux Trip Settings safety limit, hereafter called the Safety Umit. An MCPR Umit of 1.08 shall apply during single-loop operation. IRM - The IRM flux scram setting shall be set at l

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<120/125 of full scale.

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1.1 BASES 1.1 FUEL CLADDING INTEGRITY A.- Reactor Pmssum ' M @ W Ne h MM N Onset of transition boiling results in a decrease in heat transfer The fuel cladding integrity limit is set such that no calculated from the clad and, therefore, elevated clad temperature and the '

fuel damage would occur as a result of an abnormal possibility of clad failure. However, the existence of cntcal operational transient.' Because fuel damage is not directly .

observable, a step-back approach is used to establish a Safety power, or boiling transition,is not a directly observable Umit such that the minimum critical power ratio (MCPR) is no parameter in an Opmeting reactor. Therefore, the margin to less than 1.07. MCPR > 1.07 represents a conservative margin boiling transstion is calculated from plant operating parameters l such as core power, core flow, feedwater temperature, and -

relative to the conditions required to maintain fuel cladding core power distribution. The margin for each fuel assembly is integrity. The fuel cladding is one of the physical barriers which characterized by the critical power ratio (CPR) which is the separate radioactive materials from the environs. The integnty ratio of the bundle power which would produce onset of of this cladding barrier is related to its relative freedom from transition boiling dmded by the actual bundle power. The perforations or cracking. Although some corrosion or use minimum value of this ratio for any bundle in the core is the related cracking may occur during the life of the cladding, minimum critical power ratio (MCPR). It is assumed that the fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding plant operation is controlled to the reinnai protective setpoints -

via the instrumented variable, i.e., the operating domain. The perforations, however, can result from thermal stresses which current load line limit analysis contains the current operating occur from reactor operation significantly above design domain map. The Safety Umit (MCPR of 1.07) has sufficient conditions and the protection system safety settings. While fission product migration from cladding perforation is just as - conservatism to assure that in the event of an isrema! '

measurable as that from use related cracking, the thermally - operational transient initiated from the MCPR operating conditions in specification 3.1.8, more than 99.9% of the fuel caused cladding perforations signal a threshold, beyond which rods in the core are expected to avoid boiling transition. The '

still greater thermal stresses may cause gross rather than MCPR fuel cladding safety limit is increased by 0.01 for single-incremental cladding deterioration. Therefore, the fuel cladding loop operation as discussed in Reference 2.' The margin Safety Umit is defined with margin to the conditions which l.

between MCPR of 1.00 (onset of transition boiling) and the would produce onset of transition boiling, (MCPR of 1.00).

l These conditions represent a significant departure from the Safety Umit is derived from a detailed statistical analysis ]

considering all of the uncertainties in monitoring the core condition intended by design for planned operation.

operating state including the uncertainty in the boiling transition -

correlation as described in Reference 1. The uncertanties employed in deriving the Sabty Umit are Amendment No. 14,18,21,30,46,7f,98,Jd 12

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. ATTACHMENT 11

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SAFETY EVALUATION FOR PROPOSED TECH 6GES REGARDING SAFETY UMIT MINIMUM CRITICAL POWER RATIO JPTS-89 036

'k New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333

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Page 1 of 3 -

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- 12 DESCRIPTION OF THE PROPOSED CHANGES T

The proposed changes to the James A. FitzPatrick Technical Specifications revise-Specification 1.1.A and the associated Bases on pages 7 and 12.

Page 7, Specification 1.1.A Replace "1.04" with *1.07" and replace "1.05" with '1.08.*

Page 12, Bases 1.1 In two places, replace *1.04' with '1.07*

Replace "1.0" with "1.00.*

Page 12,- Bases 1.1.A g Replace "1.04* with "1.07."

Replace "1.0" with "1.00."

11. -- PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed Technical Specification changes is to revise the safety limit Minimum Critical Power Ratio (MCPR) to the appropriate value to support the Reload

, 9/ Cycle 10 core. Reload 9 will consist nf General Electric GE-10 fuel assemblies

. designated GE8x8NB 3 in GE correspondence with the NRC and four GE 11 Lead Test Assemblies. The revised safety limit MCPR will be conservatively applied to all the fuel in the core. The MCPR value of 1.0 in the Bases is modified to 1.00 to show the same number of signficant figures as the safety limit MCPR.

These fuel assemblies have two unique features: an interactive channel design and an offset lower tie plate. The offset lower tio plate shifts the fuel bundle 40 mils towards the

' control blado, making the D-lattice core of the FitzPatrick plant more like the C-lattice cores of later BWR plants.

Although this reload core affects other cycle-specific parameter limits contained in the technical specifications, no other changes are required. The Authority is removing the remainder of the cycle-specific limits from the technical specifications in a separate amendment requcat. These other changes will be reflected in a Cycle 10 Core Operating Umits Report (COLR) in accordance with the guidance contained in Generic Letter 8816.

The COLR will be provided to the NRC prior to the startup of Cycle 10. Because the change requested in this application involves revision to a safety limit, it is inappropriate to remove it from the specifications, even though its value may change based upon the fuel design selected for the reload core (Reference 3).

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. SAFETY EVALUATION Page 2 of 3

- Ill. IMPACT OF THE PROPOSED CHANGES -

The unique features of the GE8x8NB 3 fuel design have been reviewed fully and approved by the NRC for use in BWRs. In the NRC acceptance of this fuel design (Reference 1), the NRC stated that the GE8x8NB C lattice safety limit MCPR is acceptable for the GE8x8NB-3 fuel design. -This safety limit of 1.07 is provided in Table 4 2, " Fuel Cladding Integrity Safety Umit MCPR" of General Electric's " Standard Application for Reactor Fuel" (Reference 2). This revised MCPR safety limit provides the same assurance as the previous safety limit value in preventing boiling transition. All aspects of this reload fuel design and its use at the FitzPatrick plant have been reviewed and approved by the NRC.

The application of this fuel design at the FitzPatrick plant allows greater operational

. flexibility and improved fuel performance, in addition, the enrichment and gadolinium loadings are designed to support the longer,24 month cycle length planned for Cycle 11.

l. IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION -

Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since It would not: ,

- 1. . Involve a s!gnificant increase in the probability or consequences of an accident previously evaluated.

L NRC approved methodologies and codes have been used to perform all analyses concerning General Electric Co. fuel to be loaded at this refueling (Reference 1). The fuel design has been reviewed and approved for use at the FitzPatrick plant under - the constraints and methodologies detailed in References 1 and 2. There are no unique aspects of this fuel or its application L which have not undergone prior NRC review and approval. The refueling of l the FitzPatrick reactor and Cycle 10 operation does not result in an increase in the probability or consequences of any accident previously evaluated.

2. create the possibility of a new or different kind of accident from any "

accident previously evaluated.

Refueling the FitzPatrick reactor Is a periodic evolution performed in g- accordance with appropriate procedures and controlled by the Technical Specifications. The GE-10 fuel bundles inserted as Reload 9 have been fully L reviewed by the NRC (Reference 1), and their use will not create the possibility l of a new or different type of accident. The nuclear characteristics of the E individual fuel bundles and the core loading pattem have been fully analyzed L by the General Electric Co. and do not create the possibility of a new or different type of accident.

!' 3. Involve a significant reduction in a margin of safety.

The analyses performed in support of this reload assure maintenance of all existing margins of safety. These analyses have resulted in core wide and

! bundle specific limits for General Electric Co. fuel which, when applied to the i

reloaded core, assure operation within the design criteria previously approved in References 1 and 2. The revised MCPR safety limit provides an equivalent margin of safety as the previous safety limit value in preventing boiling L

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i transition. Therefore, the proposed change does not result in the reduction of any margin of safety.

V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

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VI. CONCLUSION The change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, it:

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 Analysis Report;

b. will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report;

c. will not reduce the margin of safety as defined in the basis for any technical specification; and

d. involves no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. NRC letter, A. C. Thadani to J. S. Charnley (GE), " Acceptance for Referencing of Amendment 21 to General Electric Ucensing Topical Report NEDE 24011-P-A, ' General Electric Standard Application for Reactor Fuel,'" dated March

= 17,1989.

2. " General Electric Standard Application for Reactor Fuel," Revision 9, dated September,1988.

3. GE letter, J. S. Charnley to M. W. Hodges (NRC), " Acceptance implementation of Generic Letter 88-16," dated August 8,1989.

4. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report.

5. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER),

dated November 20,1972, and Supplements.

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