ML20206A550

Application for Amend to License NPF-39,changing Tech Specs to Accommodate First Refueling w/268 New Fuel Bundles

ML20206A550
Person / Time
Site:	Limerick
Issue date:	04/03/1987
From:	Gallagher J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19292H074	List: ML20206A541 ML20206A550 ML20206A559 ML20206A569 ML20206A576
References
NUDOCS 8704080070
Download: ML20206A550 (17)

Text

r 4 re BEFORE THE 1

UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of  :

Docket No. 50-352 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE NPF-39 Edward G. Bauer, Jr.

Eugene J. Bradley 2301 Market Street Philadelphia, Pennsylvania 19101 Attorneys for Philadelphia Electric Company l

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8704090070 870403 PDR ADOCK 05000352 i p PDR, 1

l BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of  :

Docket No. 50-352 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE NPP-39 i

Philadelphia Electric Company, Licensee under Facility Operating License NPF-39 for Limerick Generating Station Unit 1, hereby requests that the Technical Specifications contained in Appendix A of the Operating License be amended as indicated by a vertical bar in the margin of the attached pages iv, vi, xviii, 2-1, 2-4, 3/4 2-1, 3/4 2-7, 3/4 2-8, 3/4 2-9, 3/4 2-10, 3/4 2-10a, 3/4 2-12, and 3/4 3-60. In addition, Licensee requests the addition of new page 3/4 2-6a to allow for additional information.

Licensee has scheduled the first LGS Unit 1 refueling outage to commence May 16, 1987. The outage is currently planned for i

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seventy-four (74) days during which time the LGS Unit 1 reactor will be reloaded with 268 new fuel bundles.

Two categories of Technical Specification changes are being proposed. The first category of change identifies the operating j limits for all fuel types for LGS Cycle 2 operation. .The second category of change incorporates a change in the slope of the flow biased Average Power Range Monitor (APRM) scram and rod block setpoints. Administrative changes are also proposed to modify the i Bases on page B 2-1 to incorporate the reload safety limit Minimum Critical Power Ratio (MCPR) and to modify the Bases on pages B 2-2 through B 2-4 and B 3/4 2-1 through B 3/4 2-5 to eliminate redundant information that is subject to periodic revision due to amendment of

" General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A.

j SAFETY ASSESSMENTS:

Category I: Operating Limits for All Fuel Types for Cycle 2 Operation i

j The proposed changes on Page 2-1 revise the Safety Limit MCPR for reload core in accordance with NEDE-240ll-P-A-8, " General Electric Standard Application for Reactor Fuel", May, 1986, which is incorporated herein by reference.

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f New fuel type GE8X8EB is planned for Cycle 2 operation.

Proposed changes on Pages 3/4 2-9, 3/4 2-10 and 3/4 2-10a identify the MCPR operating limits for the fuel being used for Cycle 2 operation.

Proposed changes on Page 3/4 3-60 modify the Rod Block Monitor (RBM) setpoint to be consistent with that used in the safety analysis. An

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I a analysis of the safety considerations involved in the reactor refueling and the Cycle 2 MCPR operating limits for all fuel types is set forth in General Electric Document No. 23A5801, " Supplemental Reload Licensing Submittal for Limerick Generating Station Unit 1, Reload 1", dated February, 1987, which is filed herewith and incorporated herein by reference. The proposed changes to the constants used to calculate the adjusted analysis mean scram time on Page 3/4 2-8 conform to Amendment 11 to the General Electric Standard Application for Reactor Fuel and were approved by the NRC in the letter from G.C. Lainas (NRC) to J.S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report, NEDE-240ll-P-A, 'GE Generic Licensing Reload Report', Supplement to Amendment 11", dated March 22, 1986.

The proposed changes incorporate the limiting power density of 14.4 kw/ft for GE8X8EB fuel on Page 3/4 2-12 and curves of the most and least limiting Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) values versus planar average exposure for GE8X8EB fuel on Pages 3/4 2-1 and 3/4 2-6a. The results of an analysis of the safety considerations involved in these MAPLHGR limits is set forth in: (1)

General Electric Company Document No. 23A5801, previously referenced, and (2) General Electric Company Document No. NEDO-31401-P, " Basis of MAPLHGR Technical Specifications for Limerick Unit 1", dated February, 1987 (which provides the specific MAPLHGR limits), which is filed l

l herewith and incorporated herein by reference. This latter document also contains MAPLHGR limits for all Cycle 1 fuel bundles and bundle descriptions for the GE8X8EB fuel to be loaded for Cycle 2. This document contains information which General Electric Company wishes to I

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maintain in confidence and to be withheld from public disclosure.

This information has been handled and classified as proprietary to General Electric Company, as indicated by the attached affidavit.

Therefore, Licensee hereby requests that this document be withheld from public disclosure in accordance with the provisions of 10 CFR i

2.790.

SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION FOR CATEGORY I:

The NRC has provided guidance concerning the application of the standards for determining whether license amendments involve no significant hazards considerations. An example of a change that involves no significant hazards consideration is a change relating to a nuclear reactor core reloading, if there are no significant changes to the acceptance criteria for the technical specifications; the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed; and the {

NRC has previously found such methods acceptable. The proposed revisions incorporate changes to the operating limits resulting from Reload 1, and the new fuel assemblies have already been found acceptable by the NRC. The NRC has previously approved GE8X8EB fuel generically as documented by letter, C.O. Thomas (NRC) to J.S.

Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-240ll-P-A-6, Amendment 10, ' General Electric Standard Application for Reactor Fuel'", dated May 28, 1985 and by letter, H.N. Berkow (NRC) to J.S. Charnley (GE), " Acceptance for Ap;iroval of Fuel Designs Described in Licensing Topical Report NEDE-240ll-P-A-6, Amendment 10 for Extended Burnup Operation", dated December 3, 1985. The NRC has

i also previously approved the application of GEMINI analysis methods, used in Document No. 23A5801, to reload licensing as documented by letter, G.C. Lainas (NRC) to J.S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-240ll-P-A, 'GE Generic Licensing Reload Report', Supplement to Amendment 11", dated March 22, 1986.

A. SAFETY LIMIT MCPR FOR RELOAD CORES

1. The safety limit MCPR is set such that no fuel damage is calculated to occur if the limit is not violated. It is determined using the NRC approved General Electric Thermal Analysis Basis (GETAB), which is a statistical model that combines uncertainties in operating parameters with uncertainties in the methods used to calculate critical power.

For reload cores, some of the uncertainties used in the determination of the safety limit MCPR are larger than for initial cores. The higher safety limit MCPR of 1.07 for reload cores accounts for these increased uncertainties. The safety limit MCPR for reload cores has received previous NRC approval and is documented in NEDE-24011-P-A-8, previously referenced.

Because the new safety limit MCPR is set such that no fuel damage is calculated to occur and thereby accomplishes the same purpose as the previous limit, this change does not l

increase the probability or consequences of any accident previously evaluated.

2. The safety limit !.JPR is not the initiating event of any accident and requires no changes in the current mode of operation and so this change does not create the possibility of a new or different kind of accident than previously evaluated.

3. Increasing the safety limit MCPR maintains the margin of safety established by the current limit. Both limits have received previous NRC approval in NEDE-240ll-P-A-8. This change does not involve a significant reduction in the margin of safety.

B. GE8X8EB MAPLHGRs

1. 10CFR50.46 establishes acceptance criteria for fuel and Emergency Core Cooling Systems-(ECCS). MAPLHGR limits are established to ensure that the acceptance criteria are met.

This change provides MAPLHGR limits for the BC320A (GE8X8EB) fuel assembly. The MAPLHGRs have been calculated using NRC approved methods (NEDE-240ll-P-A-8, previously referenced),

and the results of the analysis (General Electric Company i

Document No. 23A5801, previously referenced) demonstrate that i the acceptance criteria of 10CFR50.46 are met with substantial margin.

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This change therefore does not increase the probability or consequences of an accident previously evaluated (LOCA).

2. MAPLHGR limits are not the initiating event of any accident ,

and they do not require any changes to the current mode of operation, and so this change does not create the possibility of a new or different kind of accident than previously evaluated.

3. The ' acceptance criteria of 10CFR50.46 establisit the margins of safety for fuel and the ECCS. Calculations using NRC approved methods described in NEDE-240ll-P-A-8, previously referenced, yield results well within these acceptance criteria. The maximum peak cladding temperature (PCT) for the BC320A fuel assembly is 2055 degrees F providing 145 degrees F margin to the 2200 F limit. The maximum PCT in Cycle 1 was 2090 degrees F for BP8X8R fuel. Therefore, the proposed amendment does not result in a reduction in the margin of safety.

C. Constants Used to Determine SI

1. The constants used to calculate the adjusted analysis mean scram timeif, are the mean and standard deviation of the scram speed data base used in the determination of NRC approved MCPR adjustment factors. These new constants result in the calculation of a smaller and therefore more l

restrictive 7'. Plants unable to meet the T' criterion must 1

increase their operating limit MCPR. Plants that meet the l

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1Deriterion operate with a MCPR limit defined in the Technical Specifications.

In summary, the MCPR operating limit is set such that fuel damage is not calculated to occur during a transient. The criterion is applied to verify that the limit is appropriate.

Since the new constants result in a more restrictive t' ,

this change does not involve an increase in the probability or consequences of an accident previously evaluated.

2. The definition of 17 is not the initiating event of any accident and does not require any changes to the current mode of operation, and so this change does not create the-possibility of a new or different kind of accident than previously evaluated.

3. The change in the constants used to calculate T' results in a more restrictive criterion for applying the appropriate MCPR operating limit. The margin of safety is provided by the MCPR limit, which is increased if the 1Tcriterion is not met. Therefore, this change does not result in a reduction in the margin of safety.

D. LHGR Limit for GE8X8EB Fuel

1. In NUREG-0800, " Standard Review Plan for 4.2, Fuel System Design, Revision 2, July, 1981, the NRC provides acceptance criteria for fuel designs under accident conditions.

Evaluations have been performed of the GE8X8EB design with i ,

the higher (14.4 kw/ft) LHGR limit using NRC approved methods described in NEDE-240ll-P-A-8, previously referenced. The results of these evaluations are documented in NEDE-240ll-P-A-8, previously referenced, and demonstrate that the NRC acceptance criteria are met with the higher LHGR limit.

Because the acceptance criteria in NUREG-800 are all met, the proposed change does not increase the probability or consequences of an accident previously evaluated.

2. The LHGR limit is not the initiating event of any accident, and does not require any changes in the mode of operation, so the possibility of a new or different kind of accident than previously evaluated is not created.

3. The margin of safety is defined by the NRC acceptance criteria of NUREG-0800. The NRC has reviewed and approved General Electric evaluations demonstrating that the GE8X8EB design is within the acceptance criteria with a LHGR of 14.4 kw/ft. Therefore, the proposed amendment does not involve a reduction in the margin of safety.

E. New Operating Limit MCPRs

1. The operating limit MCPR is set such that the safety limit MCPR cannot be violated in the unlikely event of a transient.

The operating limit MCPR is determined by adding the delta-CPR for the worst transient to the safety limit MCPR. The transient delta-CPR is calculated using NRC approved methods l I

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described in NEDE-240ll-P-A-8, previously referenced.. The limiting abnormal operational transients have been re-evaluated in detail for standard conditions: generator load rejection with bypass failure, loss of 100 degrees F feedwater heating and feedwater controller failure. The limiting abnormal operational transient, feedwater controller failure, was re-evaluated in detail for operation in the Extended Load Line Limit analysis region and for operation at Increased Core Flow with or without Final Feedwater Temperature Reduction.

The Cycle 2 operating limits are based on the results given in General Electric Document No. 23A5801, previously referenced. Operation above these limits means that no fuel damage is calculated to occur in the event of a transient.

Since the Cycle 2 operating limits accomplish the same purpose as the previous limits, this change does not increase the probability or consequences of an accident previously evaluated.

2. The operating limit MCPR is not the initiating event of any accident and it does not require a change in the current mode-of operation, so the possibility of a new or different kind of accident than previously evaluated is not created.

3. The operating limit MCPR is set to prevent calculated fuel damage during the worst transient. The delta-CPR for the worst transient is determined using NRC approved methods and procedures described in NEDE-240ll-P-A-8, previously

referenced. Margin is included in the delta-CPR for the worst transient by assuming conservative input parameters and by conservatively treating uncertainties. Margin is also included in the determination of the safety limit MCPR.

Changing the operating limit MCPR does not reduce the margin of safety included in the delta-CPR and safety limit MCPR calculations.

F. Change in RBM Setpoint

1. The purpose of the Rod Block Monitor (RBM) is to prevent withdrawal of a control tod beyond the point where fuel damage is expected to occur. Evaluations have been performed using NRC approved procedures and methods (NEDE-240ll-P-A-8, previously referenced) to demonstrate the adequacy of the RBM setpoint. The results of this evaluation are provided in General Electric Document No. 23A5801, previously referenced.

The delta-CPR associated with withdrawing the maximum reactivity control rod during a Rod Withdrawal Error, to various RBM setpoints is determined. The MCPR limit for this event is determined by adding the delta-CPR at the setpoint of interest (107%) to the safety limit MCPR. Operation of the plant above this limit means that, in the unlikely event of a rod withdrawal to its maximum allowed by the RBM setpoint, fuel damage will not be calculated to occur.

The operating limit for Cycle 2 (justified in Section V of 23A5801) precludes the possibility of calculated fuel damage during all transients, including Rod Withdrawal Error with a RBM setpoint of 107%. Therefore, this change does not increase the probability or consequences of an accident previously evaluated.

2. Adjusting the RBM setpoint impacts only the Rod Withdrawal Error, which is analyzed in General Electric Document No.

23A5801, so the possibility of a new or different kind of accident is not created.

3. The delta-CPR for rod withdrawal to the RBM setpoint of 107%

is calculated using NRC approved procedures and methods (NEDE-240ll-P-A-8). Margin is included in the delta-CPR by the conservative manner in which it is calculated. Margin is also included in the determination of the safety limit MCPR.

Changing the RBM setpoint does not reduce the margin of safety included in these calculations. The Rod Withdrawal Error transient is bounded by the Feedwater Controller Failure as shown in General Electric Document No. 23A5801.

SAFETY ASSESSMENT:

Category II: Flow Biases APRM Scram and Rod Block Setpoints The proposed changes incorporate a change in the slope of the flow biased APRM scram and rod block setpoints from 0.66W to 0.58W as shown on pages 2-4, 3/4 2-7 and 3/4 3-60. An analysis of the safety considerations involved in this change is set forth in General Electric Document No. 23A5801, previously referenced. This safety analysis refers to analyses to support operation in the extended load line limit analysis region set forth in General Electric Document NEDC-31139, " General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Limerick Generating Station, Unit 1, Cycle 1",

dated April 1986, which is filed herewith and incorporated herein by reference.

SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION IN CATEGORY II:

The NRC has provided guidance concerning the application of the standards for determining whether license amendments involve no significant hazards considerations by providing examples (51FR7751).

An example (iv) of a change that involves no significant hazards consideration "a relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated." These proposed revisions fit this example since the effect of the change in slope of the flow biased APRM scram and rod block setpoints is to provide slightly increased extended load line analysis region, operation in which is justified by analyses documented in NEDC-31139 and confirmed in Document No. 23A5801.

Change in Slopes of APRM Scram and Rod Block

1. The change in APRM scram and rod block setpoint equations was evaluated using NRC approved procedures and methods. The results of this evaluation are demonstrated in NEDC-31139, previously referenced. Application of this change in APRM scram and rod

block setpoint equations to Cycle 2 is confirmed in General Electric Document No. 23A5801, previously referenced.

In NEDC-31139, the impact of the new APRM scram and rod block equations on overpressure protection, stability, loss of coolant accident, containment, reactor internals, and anticipated transients without scram (ATWS) events was evaluated. The results of these evaluations demonstrate that all design limits identified in the FSAR are met.

Because operation with the APRM scram and rod block setpoint equation is well within the bases reviewed and approved by the NRC in the FSAR, this change does not increase the possibility or consequences of an accident previously evaluated.

2. The impact of changing the APRM scram and rod block setpoint equations has been considered for all transients and accidents identified in the FSAR. The possibility of a new or different kind of accident from any accident previously evaluated is not created.

3. NEDC-31139 demonstrates that the change in slope of the APRM scram and rod block setpoint equations is within the limits identified in the FSAR. General Electric Document No. 23A5801 confirms that the operating limit MCPRs determined in General Electric Document No. 23AS801 bound operation with the new equations. Thus, the margin of safety is not reduced.

Environmental Consideration Proposed amendment allows Limerick Unit 1 to operate under the identified operating limits for all Cycle 2 fuel types. It further allows Cycle 2 operation with a change in the slope of the flow biased APRM scram and rod bloc,k setpoints. Licensee has determinedthattheproposedamendmentinholvesnoincreaseinthe amounts and no change in the types of effluents that may be released offsite and has also determined that there is no increase in the individual or cumulative occupational radiation exposure.

The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Technical Specifications, and have concluded that they do not involve unreviewed safety questions or involve a significant hazards consideration and will not endanger the health and safety of the public.

Respectfully submitted, PHILADELPHIA ELECTRIC COMPANY By CPice PresFdent

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l COMMONWEALTH OF PENNSYLVANIA :

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l COUNTY OF PHILADELPHIA  :

J. W. Gallagher, being first duly sworn, deposes and says: -

That he is Vict President of Philadelphia Electric Company, the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating License and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

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Subscribed and sworn to before me this.3 ay af QALN, /1V7 2_e

, ,e Notary Public MELANIE R. CAMPANELfA Not6ry Pubik, PMedelphia. Philadelphia Co.

My Commluma Espres rebrusty 12,1990 i

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