ML20245G558

Application for Amends to Licenses DPR-57 & NPF-5,revising Tech Spec for Thermal Power & Power Distribution Limits (Fuel Rods),Reactivity Control Sys & Definitions & Administrative Controls

ML20245G558
Person / Time
Site:	Hatch
Issue date:	06/22/1989
From:	Hairston W GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20245G563	List: ML20245G558 ML20245G565
References
NUDOCS 8906290178
Download: ML20245G558 (14)

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June 22, 1989 U. S. Nuclear Regulatory Commission ATTN: Document Control Room Washington, D. C. 20555 EDWIN 1. HATCH NUCLEAR PLANT - UNITS 1 AND 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS:

REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS Gentlemen:

In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company (GPC) hereby proposes a change to the Plant Hatch Units 1 and 2 Technical Specifications, Appendix A to Operating Licenses DPR-57 and NPF-5.

The proposed changes involve revisions of the Plant Hatch Units 1 and 2 Technical Specifications for Thermal Power, Power Distribution Limits (Fuel Rods), Reactivity Control Systems, and Definitions and Administrative Controls.

Specifically, the proposed Technical Specifications changes will:

1. Modify the core-related specifications having cycle-specific parameter limits by replacing the values of those limits with a reference to a Core Operating Limits Report which will contain the values of those limits. The proposed changes include the addition of the Core Operating Limits Report to the Definitions  !

section and the addition of reporting requirements to the Administrative Controls section of the Technical Specifications.

These changes are consistent with the guidance for making such changes that the NRC provided in Generic Letter 88-16.

2. Reduce the core power level below which Control Rod Program Control operation is required from 20 percent to 10 percent of rated thermal power.

3. Revise the Bases and Definitions to permit use of NRC-approved transition boiling correlations other than GEXL.

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U.S. Nuclear Regulatory Commission Page 2 June'22, 1989 .

These changes will reduce the time and effort required by the NRC and GPC to process . Technical Specifications changes related to cycle-specific parameters. These charqes will also permit' improvements in plant capacity factor by reducing the power range in which Group Notch control rod movement must be enforced.

Enclosure 1 provides a detailed description of the proposed change and the. circumstances necessitating the' change request.

Enclosure 2 details the bases for our determination that the proposed change does not involve a significant hazards consideration.

Enclosure 3 provides page change instructions for incorporating the proposed change. The proposed changed Technical Specifications pages for Unit 1 and Unit 2 follow Enclosure 3.

l To allow time for procedure revisions and orderly incorporation into L

copies of the Technical Specifications, GPC requests the proposed amendment, once approved by the NRC, be issued with ar. effective date to be no later than 60 days from~the date of issuance of the amendment. In accordance with Generic Letter 88-16 and the proposed Administrative Control reporting requirement, GPC will provide the NRC a Core Operating Limits Report for each Hatch unit at the time the limits in that report become effective; transmittal of the Core Operating . Limits Report will be no later than 60 days following NRC approval of this amendment request.

For reactor cycles that will contain axially-zoned fuel assemblies which have been generically approved by the NRC, GPC will provide the NRC the fuel assembly's nuclear design information, if that information has not been previously submitted by GPC or the fuel vendor. As discussed in Section 2.8 of the NRC safety evaluation for Amendment 18 to NEDE-240ll-P-A, "GE Standard Application for Reactor Fuel (GESTAR-II)",

, the nuclear design information for such fuel is. for NRC information only; l therefore, the design information will be submitted on the same schedule as the Core Operating Limits Report.

In accordance with the requirements of 10 CFR 50.91, a copy of this letter and all applicable enclosures will be sent to Mr. J. L. Ledbetter of the Environmental Protection Division of the Georgia Department of Natural Resources.

Based on assurances contained in Generic Letter 88-16, GPC requests that the proposed amendment be expeditiously reviewed and issued as soon l as possible.

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~ U.S. Nuclear Regulatory Commission Page 3 June 22, 1989 Mr. W. G. Hairston, III states that he is duly authorized to execute this oath on behalf of Georgia Power Company, and to the best of his knowledge and belief, the facts set forth in this letter are true.

GEORGIA POWER COMPANY By: Ld YNC W. G. Hairston, III Sworn to and subscribed before me this JJ day of June 1989.

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Enclosures:

1. . Basis for Change Request

2. 10 CFR 50.92 Evaluation

3. Page Change Instructions cc: Georaia Power Company Mr. J. T. Beckham, Jr., Vice President - Plant Hatch Mr. L. T.- Gucwa, Manager, Engineering and Licensing - Plant Hatch GO-NORMS ,

U.S.-Nuclear Reaulatory Commission. Washinaton. D. C.

' Mr. L. P. Crocker Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaic, II Mr. S. D. Ebneter, Regional Administrator

.Mr. J. E. Menning, Senior Resident Inspector - Hatch State of Georoia Mr. J. L. Ledbetter, Commissioner - Department of Natural Resources j l

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c ENCLOSURE 1 i

PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS BASIS FOR CHANGE RE0 VEST PROPOSED CHANGE 1:

This change is proposed to the Technical Specifications for Edwin I. Hatch Nuclear Plant Units 1 and 2. This proposed change would modify specifications having cycle-specific parameter limits by replacing the values of those limits with a reference to a Core Operating Limits Report which includes the cycle-specific val ues . This proposed change also includes the addition of a Core Operating Limits Report to the Definitions section and the addition of reporting requirements to the Administrative Controls section of the Technical Specifications.

BASIS FOR PROPOSED CHANGE 1:

A portion of this change is administrative in nature since information presently in the Technical Specifications is simply being relocated to a Core Operating Limits Report. This information includes cycle-specific power distribution limits (e.g., Minimum Critical Power Ratio (MCPR) and Average Planar Linear Heat Generation Rate (APLHGR)], which are calculated using NRC-approved methodology. The change will provide additional margin to fuel thermal limits by allowing the Minimum Critical Power Ratio (MCPR) Operating Limits to be set based on cycle-specific core designs and analyses, instead of non-cycle-specific, bounding values.

Guidance on this change was developed by the NRC on the basis of the review of a lead-plant proposal submitted on the Oconee plant docket that was endorsed by the Babcock and Wilcox Owners Group. This guidance was provided to all power reactor licensees and applicants by Generic Letter 88-16 dated October 4, 1988. This proposed change to the Technical Specifications is fully consistent with that guidance.

According to Generic Letter 88-16, the alternative to including the values of cycle-specific parameter limits in individual specifications includes (1) the addition of a newly defined term for the formal report that provides the cycle-specific parameter limits, (2) the addition of its associated reporting requirement to the Administrative Controls section of the Technical Specifications, and (3) the modification of individual specifications to replace these limits with a reference to the defined formal report for the values of these limits.

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ENCLOSURE 1

RE00EST TO REVISE TECHNICAL SPECIFICATIONS

REACTIVITY CONTROL SYSTEMS AND l REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS 1

BASIS FOR CHANGE RE0 VEST .,

k L This proposed change will modify the Definitions section of the Technical Specification, to include a definition of the Core Operating Limits Report 1 that documents cycle / reload-specific parameter limits based on NRC-approved )

methodology. The definition notes that plant operation within these limits i is addressed by individual specifications. 8 The following specifications are being revised to replace the values of  !

cycle-specific parameter limits with a reference to a Core Operating Limits Report that provides these limits: { 4 Unit 1 3.3.F Operation With a Limiting Control Rod Pattern (for Rod .

Withdrawal Error, RWE) l 3.11.A Average Planar. Linear Heat Generation Rate (APLHGR) 3.11.C Minimum Critical Power Ratio (MCPR) .i 4.11.C Minimum Critical Power Ratio (MCPR)

Unit 2 3.1.4.3 Rod Block Monitor 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.3 Minimum Critical Power Ratio The Technical Specifications Bases for the sections listed above will be '

modified as necessary to refer to the Core Operating Limits Report for the cycle-specific limits, instead of providing specif?.c values or referring to Technical Specifications figures.

Specification 6.9.1.11 is being added to the reporting requirements of the Administrative Controls section of the Technical Specifications. This specification requires that the Core Operating Limits Report be submitted, upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. The report provides the values of cycle-specific parameter limits that are applicable for the current fuel cycle. Furthermore, this specification requires that the i

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ENCLOSURE 1 RE0 VEST TO PEVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS _

BASIS FOR CHANGE RE0 VEST values of these limits be established using the NRC-approved methodology described in NEDE-240ll-P-A and in the NRC Safety Evaluations on the Hatch ANF 9x9 Lead Fuel Assemblies. The values of the cycle-specific limits will be established consistent with all applicable limits of the safety analysis. Finally, the specification requires that all changes in cycle-specific parameter limits be documented in the Core Operating Limits Report before each reload cycle or remaining part of a reload cycle, and submitted upon issuance to the NRC prior to operation with the new parameter limits. q Because plant operation will continue to be limited in accordance with the values of cycle-specific parameter limits that are established using an NRC-approved methodology, this change is administrative in nature.

Therefore, there will be no impact on plant safety.

PROPOSED CHANGE 2:

The Control Rod Program Controls (CRPCs) at Hatch Units 1 and 2 consist of the Rod Sequence Control System (RSCS) and the Rod Worth Minimizer (RWM).

This change will reduce the core power level below which operation of a CRPC is required from 20 percent to 10 percent of rated thermal power.

The change will also revise Hatch Unit 2 Technical Specifications for other plant systems which refer to this power level to maintain unchanged operability and surveillance requirements for those systems. A change is proposed to clarify the intent of the Unit 1 RSCS surveillance requirement, and appropriate changes ~ are proposed for the Technical Specifications Bases. ,

BASIS FOR PROPOSED CHANGE 2:

Amendment 17 to NEDE-240ll-P-A (GESTAR-II) justified reduction of the power level at which the CRPCs are bypassed from its current value of 20 percent to 10 percent of rated thermal power. That justification was based on the fact that the analytical basis for this bypass power level is 10 percent. The NRC Safety Evaluation Report for Amendment 17 states that the 20-percent Technical Specification limit was previously required as an extreme bound, because of uncertainties in the Rod Drop Accident (RDA) analyses available in the early 1970's. It is now recognized that if the core power level exceeds 10 percent of rated thermal power, no control rod pattern can generate rod worths such that the fuel enthalpy would exceed the 280 cal /gm fuel enthalpy limit during the worst RDA. For HL-246 44200 El-3

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ENCLOSURE I RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS BASIS FOR CHANGE RE0 VEST I

this reason, this reduction in the bypass power level. analytical limit (Technical -Specifications) value was approved by the NRC through its approval of Amendment 17.

The Hatch Technical Specifications now require that the RSCS be operable i and that Group Notch rod movement be followed while control rod density is below 50 percent'and power is below the bypass power level. (Group Notch movement requires individual control rods in a given RSCS group to be moved, a single notch at a time, in a round-robin fashion.) Group Notch movement is very time-consuming, and is associated with poor human factors for the reactor operator. The Technical Specifications also require that whenever the reactor is in the Start & Hot Standby or Run Mode below 20 l percent of rated thermal power, the RWM shall be operable or a second licensed operator shall verify that the operator at the reactor console is i following the control rod program. Therefore, reducing the power level at which the RSCS and RWM are automatically bypassed will simplify plant operation, reduce startup-shutdown times, and permit quicker power reductions in emergency conditions.

Hatch Unit 2 Technical Specifications for Control Rod Operability (4.1.3.1), Rod Block Monitor (RBM) (3.1.4.3 and 3/4.3.5), and Special Test Exceptions (4.10.2) currently refer to the preset power level of the RWM and/or RSCS. The preset power level corresponds to the instrument setpoint at which the RWM and RSCS are automatically bypassed; the current nominal value of the preset power level is 30 percent of rated thermal power.

The change proposed for Unit 2 Technical Specification 4.1.3.1 will maintain essentially unchanged Surveillance Requirements after the bypass power is reduced. The weekly control rod surveillance will be required for power levels above 30 percent of rated. In addition, the power level above which RBM operability is required will be changed in Technical Specifications 3.1.4.3 from "the preset power level of the RWM and RSCS" -

to' "30 percent of RATED THERMAL POWER." Specification 3/4.3.5 will be I changed to delete the reference to the preset power level. The reference to 3.1.4.3, where the power level requirement is provided, will be retained. These changes are consistent with the existing Low Power i Setpoint of the RBM and with the RBM design basis. Finally, the reference to the RSCS preset power level in Technical Specification 4.10.2 will be l

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ENCLOSURE 1 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS BASIS FOR CHANGE RE0 VEST replaced with "10% of RATED THERMAL POWER," because that Specification refers to the required power range for RSCS operability, which is being changed.

The existing Unit 1 Technical Specification .4.3.G.2.a for RSCS surveillance requires testing of the RSCS rod inhibit function under-certain conditions. The proposed change to this Specification will clarify the intent of this requirement and make the Unit I requirement consistent with Unit 2.

It should be noted that this change eliminates a potential. conflict between maintaining proper rod pattern control and following our current operating practices relative to thermal-hydraulic stability.' NRC Bulletin 88-07, Supplement 1, " Power Oscillations in BWRs," was issued on December 30, 1988. The Bulletin Supplement provided a set of operating recommendations to reduce the likelihood of a thermal-hydraulic insta'ility event. If certain regions of power-to-flow operating map are entered, the operator is instructed to insert a predefined set of high worth or " cram" control rods to rapidly reduce core thermal power. One potential conflict has been identified by the BWR Owner's Group where, during cram rod insertion, the power level could drop below the actual instrument low power setpoint (LPSP), causing the RWM and/or RSCS to " lock up" control rods. Lowering the analytical LPSP to 10% power would make the actual plant setpoint low enough, so " cram" rod insertion would not be expected to be affected by RSCS/RWM operation.

PROPOSED CHANGE 3:

This change will revise the Technical Specifications Definitions (Unit 2) and Bases (Units 1 and 2) replacing references to the GEXL transition i boiling correlation with references to "an NRC-approved critical power (or transition boiling) correlation." This change is intended to permit the use of approved correlations other than GEXL for determination of the Minimum Critical Power Ratio (MCPR).

BASIS FOR PROPOSED CHANGE 3:

The NRC recently approved two new transition boiling correlations, referred to as GEXL-PLUS and GEXL-PLUS GE8x8NB, for use in determining the  !

MCPR for GE fuel. When certain conditions apply, these correlations may be used for reload safety analysis and for core monitoring.

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ENCLOSURE 1 l RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS i REACTIVITY CONTROL SYSTEMS AND  ?

REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS BASIS FOR CHANGE RE0 VEST Because the transition boiling correlation to be used may differ as fuel changes from cycle to cycle, the specification of a particular correlation ,

to be used is effectively a cycle-specific parameter. Therefore, it is '

proposed that each reference to GEXL in the Unit I and Unit 2 Technical Specifications be replaced with a reference to "an NRC-approved critical power (or transition boiling) correlation." This will. permit GPC to use the appropriate, approved correlation for each reload fuel type without amending the Technical Specifications. All changes to the MCPR margin

, required for conservative application of the approved correlations will be-  !

accommodated either through adjustments to the operating MCPR calculated by the plant process computer or to the MCPR limits to be provided in the i Core Operating Limits Report.

GPC proposes to insert references - to the generic GE reload fuel topical report, NEDE-24011-P-A, into the Technical Specifications Bases to replace information on the development and application of GEXL that may not apply

_to the other NRC-approved transition boiling correlations. Similar l information for each of the approved transition boiling correlations is presented in NEDE-24011-P-A.

This proposed change will introduce flexibility to apply approved boiling correlations that GE has developed to be compatible with their approved fuel designs; it will not involve any changes in fuel design and plant operation will not be significantly affected. .Use of other approved

' correlations may facilitate . operation with more efficient fuel-loading configurations; however, the NRC-approved methods will. continue to be applied to demonstrate compliance of each Hatch core to all Specified Acceptable Fuel Design Limits.

l Therefore, this change is purely administrative and will have no impact on plant safety.

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ENCLOSURE 2 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS 10 CFR 50.92 EVALUATION PROPOSED CHANGE 1:

Description of Amendment Reauest: The proposed amendment would revise Technical Specifications 3.3.F, 3.11.A, 3.11.C, 4.11.C, and associated Bases of the Unit 1 Technical Specifications and 3.1.4.3, 3/4.2.1, 3/4.2.3, and associated Bases of the Unit 2 Technical Specifications to replace the values of cycle-specific parameter limits with a reference to the Core Operating Limits Report which contains the values of those l limits. In addition, the Core Operating Limits Report has been included in the Definitions Section of the Technical Specifications to note that it is the unit-specific document that provides these limits for the current operating reload cycle. Furthermore, the definition notes that the values of these cycle-specific parameter limits are to be determined in accordance with Specification 6.9.1.11. This Specification requires that the Core Operating Limits be determined for each reload cycle in accordance with the referenced NRC-approved methodology for these limits and consistent with the applicable limits of the safety analysis.

Finally, this report and any mid-cycle revisions shall be provided to the NRC upon issuance. Generic Letter 88-16 dated October 4, 1988, from the NRC provided guidance to licensees on requests for removal of the values of cycle-specific parameter limits from the Technical Specifications.

This proposed change is in response to Generic Letter 88-16.

Basis for Proposed No Significant Hazards Consideration Determination:

Georgia Power Company has evaluated this proposed amendment and i determined that it involves no significant hazards considerations.  !

The proposed revision is in accordance with the guidance provided in Generic Letter 88-16 for licensees requesting removal of the values of cycle-specific parameter limits from the Technical Specifications.

This proposed change does not involve a significant hazards consideration for the following reasons:

1. The proposed change on the removal of the values of cycle-specific limits does not involve a significant increase in the probability or consequences of an accident previously evaluated. Establishment of  ;

these limits in accordance with an NRC-approved methodology and the )

incorporation of these limits into the Core Operating Limits Report i i

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ENCLOSURE 2 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND  ;

REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS 10 CFR 50.92 EVALUATION I

will ensura that proper steps will be taken to establish the values of f these limits. . Furthermore, the submittal of the Core Operating Limits '

Report will allow the NRC to continue to trend the values of these limits.

2. The revised specifications with the removal of the values of ,

cycle-specific parameter limits and the addition of the referenced j report .for these limits does not create the possibility of a new or i different kind of accident from those previously evaluated.

3. The revised specifications also do not involve a reduction in the ma d n of safety, since this change does not alter the methods used to { .

establish the cycle-specific parameter limits.

Because the values of cycle-specific parameter limits will continue to be determined in'accordance with an NRC-approved methodology and consistent with the ' applicable limits of the safety analysis, this change is administrative in nature and does not impact the operation of the facility in a manner that' involves a significant hazards consideration.

The proposed amendment does not alter the requirement that the plant be operated within the limits for cycle-specific parameters or the required remedial actions that must be taken when these limits are not met. While it is. recognized that such requirements are essential.to plant safety, the values of limits can be determined in accordance with NRC-approved methods without affecting nuclear safety. With the removal of the values from the Technical- Specifications, they will be incorporated into the Core l

Operating Limits Report that is submitted to the Commission; hence, j appropriate measures exist to control the values of these limits. This '

change is administrative in nature and does not impact the operation of the facility in a manner that involves a significant hazards ,

consideration. j Based on the preceding assessment, Georgia Power Company has determined that this proposed amendment involves no significant hazards  ;

consideration.

PROPOSED CHANGE 2:

1 Description of Amendment Reauest: This change will revise Unit 1 Technical Specifications 3.3.G and 4.3.G and Unit 2 Technical Specifications 3.1.4.1, 3/4.1.4.2, 3/4-.10.2, and associated Bases to HL-246 44200 E2-2 j J

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reduce the core power level below which operation of a Control Rod Program l Control (CRPC) is required from 20 percent to 10 percent of rated thermal l power. The change will also revise Unit 2 Technical Specifications for other plant systems which refer to this power level (i.e., Technical Specifications 4.1.3.1,3.1.4.3,3/4.3.5, and associated Bases) to clarify and maintain conservative operability or surveillance requirements for l those systems. Appropriate changes are proposed for the Technical Specifications Bases. _

l Basis for Proposed No Significant Hazards Consideration Determination:

This proposed change does not involve a significant hazards consideration for the following reasons:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because it pertains to operability requirements for equipment that blocks control rod movement (except for scrams) when conditions warrant. None of the accidents previously evaluated can be initiated as a result of any action or lack of action by that equipment.

Analyses have shown that the consequences of accidents previously analyzed, including the design basis Rod Drop Accident (RDA), would not be significantly increased as a result of this change.

2. The proposed change does not create the possibility of a new or different kind of accident from any previously analyzed, because the change in plant operation associated with ttcis change is insignificant in this respect. No accidents can be initiated as a result of any action or lack of action by the equipment involved in this change.

3. The proposed change does not involve a significant reduction in the margin of safety, because the acceptance criteria for the design basis RDA are met even when this change is implemented.

PROPOSED CHANGE 3:

Description of Amendment Reauest: This change will revise the Technical Specifications Definitions (Unit 2) and Bases (Units 1 and 2), replacing the references to the GEXL transition boiling correlation with references l

to "an NRC-approved critical power (or transition boiling) correlation."

This change is intended to permit the use of approved correlations other than GEXL for determination of the Minimum Critical Power Ratio (MCPR).  ;

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ENCLOSURE 2 RE0 VEST TO REVISE TECHNICAL SPECIFICATIONS REACTIVITY CONTROL SYSTEMS AND REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS 10 CFR 50.92 EVALUATION Basis for Pronosed No Significant Hazards Consideration Determination:

The NRC recently approved two transition boiling correlations, referred to as GEXL-PLUS and GEXL-PLUS GE8x8NB, for use in determining the MCPR for GE fuel. When certain conditions apply, these correlations may be used for reload safety analysis and for core monitoring.

Because the correlation to be used may differ as reload fuel changes from cycle to cycle, the specification of the particular correlation used is effectively a cycle-specific parameter. Therefore, it is proposed that i each reference to GEXL in the Technical Specifications be replaced with a l reference to "an NRC-approved critical power (or transition boiling) correlation." This will permit GPC to use the appropriate, approved i correlation for each reload fuel type without amending the Technical Specifications, All changes to the MCPR margin required for conservative application of the approved correlations will be accommodated either through adjustments to the operating MCPR calculated by the plant process computer or to the MCPR limits to be provided in the Core Operating Limits '

Report.

GPC proposes to insert references to the generic GE reload fuel topical report, NEDE 24011-P-A, into the Technical Specifications Bases to replace information on the development and application of GEXL that may not apply to the newer, NRC-approved transition boiling correlations. Similar information for each of the approved transition boiling correlations is presented in NEDE-24011-P-A.

This proposed change will introduce flexibility to apply approved boiling correlations that GE has developed to be compatible with their approved fuel designs. This change will not involve any changes in fuel design or plant procedures.

This proposed change does not involve a significant hazards consideration for the following reasons:

1. The proposed change does not involve a significant increase in the '

probability or consequences of an accident previously evaluated because NRC-approved methods will continue to be applied to demonstrate compliance of Hatch fuel to all Specified Acceptable Fuel Design Limits.

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2. The proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. The use of other approved correlations may facilitate operation with more efficient fuel loading configurations. However, core operation will be similar enough to current practices that no new accidents could result. l

3. The proposed change does not involve a significant reduction in the margin of safety because this change does not involve a change in any fuel safety or design limits.

This proposed change does not alter any existing NRC requirements for core analysis or monitoring. Therefore, this change is purely administrative and does not impact the operation of the facility in a manner that involves a significant hazards consideration.

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