ML040690687

Licensing Amendment, Amendment Regarding Minimum Critical Power Ratio Safety Limits and Reference Changes

ML040690687
Person / Time
Site:	Susquehanna
Issue date:	03/09/2004
From:	Richard Guzman NRC/NRR/DLPM/LPD1
To:	Shriver B Susquehanna
Guzman R, NRR/DLPM 415-1030
References
TAC MB9902
Download: ML040690687 (17)

Text

March 9, 2004 Mr. Bryce L. Shriver Senior Vice President and Chief Nuclear Officer PPL Susquehanna, LLC 769 Salem Boulevard, NUCSB3 Berwick, PA 18603-0467

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMITS AND REFERENCE CHANGES (TAC NO. MB9902)

Dear Mr. Shriver:

The Commission has issued the enclosed Amendment No. 216 to Facility Operating License No. NPF-14 for the Susquehanna Steam Electric Station, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 1, 2003, as supplemented by letters dated November 17 and December 22, 2003.

This amendment revises the values of the Safety Limit for Minimum Critical Power Ratio in TS 2.1.1.2, clarifies fuel design features in TS 4.2.1, and updates the references used to determine the core operating limits.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commissions Biweekly Federal Register Notice.

Sincerely,

/RA/

Richard V. Guzman, Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-387

Enclosures:

1. Amendment No. 216 to License No. NPF-14

2. Safety Evaluation cc w/encls: See next page

March 9, 2004 Mr. Bryce L. Shriver Senior Vice President and Chief Nuclear Officer PPL Susquehanna, LLC 769 Salem Boulevard, NUCSB3 Berwick, PA 18603-0467

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMITS AND REFERENCE CHANGES (TAC NO. MB9902)

Dear Mr. Shriver:

The Commission has issued the enclosed Amendment No. 216 to Facility Operating License No. NPF-14 for the Susquehanna Steam Electric Station, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 1, 2003, as supplemented by letters dated November 17 and December 22, 2003.

This amendment revises the values of the Safety Limit for Minimum Critical Power Ratio in TS 2.1.1.2, clarifies fuel design features in TS 4.2.1, and updates the references used to determine the core operating limits.

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commissions Biweekly Federal Register Notice.

Sincerely,

/RA/

Richard V. Guzman, Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-387

Enclosures:

1. Amendment No. 216 to License No. NPF-14

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC PDI-1 RF RLaufer FAkstulewicz ACRS ZAbdullahi RGuzman MOBrien GHill (4)

OGC CBixler, RGN-1 TBoyce BPham

• SE provided. No major changes made.

• See previous concurrence Accession No.: ML040690687 Package No.: ML TSs: ML OFFICE PDI-1/PM PDI-2/LA SRXB

• OGC*

PDI-1/SC*

NAME RGuzman MO'Brien FAkstulewicz SCole RLaufer DATE 3/9/04 3/9/04 3/8/04 SE DTD 3/1/04 3/5/04 OFFICIAL RECORD COPY

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Richard L. Anderson Vice President - Nuclear Operations PPL Susquehanna, LLC 769 Salem Blvd., NUCSB3 Berwick, PA 18603-0467 Aloysius J. Wrape, III General Manager - Nuclear Assurance PPL Susquehanna, LLC Two North Ninth Street, GENA92 Allentown, PA 18101-1179 Terry L. Harpster General Manager - Plant Support PPL Susquehanna, LLC 769 Salem Blvd., NUCSA4 Berwick, PA 18603-0467 Robert A. Saccone General Manager - Nuclear Engineering PPL Susquehanna, LLC 769 Salem Blvd., NUCSB3 Berwick, PA 18603-0467 Rocco R. Sgarro Manager - Nuclear Regulatory Affairs PPL Susquehanna, LLC Two North Ninth Street, GENA61 Allentown, PA 18101-1179 Curtis D. Markley Supervisor - Nuclear Regulatory Affairs PPL Susquehanna, LLC 769 Salem Blvd., NUCSA4 Berwick, PA 18603-0467 Michael H. Crowthers Supervising Engineer Nuclear Regulatory Affairs PPL Susquehanna, LLC Two North Ninth Street, GENA61 Allentown, PA 18101-1179 Dale F. Roth Manager - Quality Assurance PPL Susquehanna, LLC 769 Salem Blvd., NUCSB2 Berwick, PA 18603-0467 Herbert D. Woodeshick Special Office of the President PPL Susquehanna, LLC 634 Salem Blvd., SSO Berwick, PA 18603-0467 Bryan A. Snapp, Esq Assoc. General Counsel PPL Services Corporation Two North Ninth Street, GENTW3 Allentown, PA 18101-1179 Supervisor - Document Control Services PPL Susquehanna, LLC Two North Ninth Street, GENTW3 Allentown, PA 18101-1179 Richard W. Osborne Allegheny Electric Cooperative, Inc.

212 Locust Street P.O. Box 1266 Harrisburg, PA 17108-1266 Director - Bureau of Radiation Protection Pennsylvania Department of Environmental Protection P.O. Box 8469 Harrisburg, PA 17105-8469 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 35, NUCSA4 Berwick, PA 18603-0035 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406

Susquehanna Steam Electric Station, Units 1 and 2 cc:

Board of Supervisors Salem Township P.O. Box 405 Berwick, PA 18603-0035 Dr. Judith Johnsrud National Energy Committee Sierra Club 443 Orlando Avenue State College, PA 16803

PPL SUSQUEHANNA, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-387 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 216 License No. NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by PPL Susquehanna, LLC, dated July 1, 2003, as supplemented by letters dated November 17 and December 22, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 216 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PPL Susquehanna, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented upon startup following the Unit 1 thirteenth refueling and inspection outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 9, 2004

ATTACHMENT TO LICENSE AMENDMENT NO. 216 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 2.0-1 2.0-1 4.0-1 4.0-1 5.0-22 5.0-22 5.0-23 5.0-23

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 216 TO FACILITY OPERATING LICENSE NO. NPF-14 PPL SUSQUEHANNA, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 DOCKET NO. 50-387

1.0 INTRODUCTION

By application dated July 1, 2003, as supplemented by letters dated November 17 and December 22, 2003, PPL Susquehanna, LLC, (PPL, the licensee), requested changes to the Technical Specifications (TSs) for Susquehanna Steam Electric Station, Unit 1 (SSES-1). The proposed change revises the values of the Safety Limit for Minimum Critical Power Ratio (SLMCPR) in TS 2.1.1.2, clarifies fuel design features in TS 4.2.1, and updates the references used to determine the core operating limits in TS 5.6.5.b. The supplements dated November 17 and December 22, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 5, 2003 (68 FR 46245).

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements The regulatory requirements and guidance which the NRC staff considered in its review of the application are as follows:

1.

Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to the reactivity control systems. Specifically, General Design Criteria 10 (GDC-10), Reactor design, in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 states, in part, that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded.

2.

NRC Generic Letter 88-16 (GL 88-16), Removal of Cycle-Specific Parameter Limits from Technical Specifications, provides guidance on modifying cycle-specific parameter limits in the TSs.

3.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, provides guidance on the acceptability of the reactivity control systems, the reactor core and fuel system design. Specifically, Section 4.2, Fuel System Design, specifies the criteria for evaluation of fuel design limits such that there be at least 95% probability at a 95% confidence level that the hot fuel rod in the core does not experience a departure from nucleate boiling or transition condition during normal operation or anticipated operational occurrence (AOO). Section 4.4, Thermal Hydraulic Design, provides guidance on the review of thermal-hydraulic design in meeting the requirement of GDC-10 and the fuel design criteria established in Section 4.2.

3.0 TECHNICAL EVALUATION

3.1 PPL/Framatome SLMCPR Methodology Framatome is the current fuel vendor for PPLs Unit 1 Cycle 14 (U1C14). However, PPL performs the reload core design and analysis, including generating the lattice neutronic data and simulating the cycle steady state core-wide neutronic and thermal-hydraulic response. PPL is currently licensed to generate the lattice neutronic data using CASMO-3G and to model the reactor steady state core response for the cycle, using MICROBURN-B (References 6 and 8).

PPL uses the CASMO-3G/MICROBURN-B code system and provides to Framatome the cycle neutronic and thermal-hydraulic response data, the core reactivity, flow, and nodal power distribution. Framatome determines the cycle SLMCPR limit that ensures 99.9% of the fuel rods will avoid boiling transition during steady state and transient events.

The NRC staff previously approved the critical power correlations applicable to the ATRIUM-10 fuel loaded for U1C14 (Reference 5). The NRC staff also approved Framatomes analytical method for calculating the SLMCPR (Reference 7). Framatome is currently licensed to use CASMO-4, a lattice spectrum/depletion code, and MICROBURN-B2, a core simulator code, to perform neutronic and thermal-hydraulic analysis for boiling water reactors (Reference 5).

Framatome also uses the upgraded CASMO-4/MICROBURN-B2 code systems in the new POWERPLEX III core monitoring system, while the POWERPLEX II core monitoring system uses the earlier NRC-approved lattice depletion code (CASMO-3) and core simulator code (MICROBURN-B). In the current Framatome SLMCPR licensing method, the CASMO-4/MICROBURN-B2 code system is used for generating the lattice neutronic data, simulating the core cycle neutronic and thermal-hydraulic response, and monitoring the core.

The approved SLMCPR calculation methodology for both Framatome and PPL are based on compatible use of code systems both in simulating the cycle core response and in monitoring the core. This provides consistency in fuel related uncertainties in the code systems used to generate the lattice neutronic design parameters and simulating the core conditions used in the core monitoring system. The NRC approval of the code systems used to simulate the reactor core conditions for the cycle includes review of the calculational uncertainties associated with the given code system. In establishing these calculational uncertainties, PPL benchmarks the code predictions of key calculated parameters and the predicted transversing incore probe readings against statistically generated data, reactor measured data, and the fuel assembly gamma scan. The uncertainties of each key predicted and measured parameters are included in the SLMCPR calculation methodology, by statistically perturbing each parameter according to the corresponding uncertainty range. Therefore, the range of the power distribution uncertainties affect the calculated SLMCPR limit for each cycle. A code system with a lower power distribution uncertainty, such as CASMO-4/MICROBURN-B2, would result in a smaller power distribution perturbation, contributing to a lower SLMCPR value for the cycle in comparison with a code system with a higher power distribution uncertainty. Specifically, Table 9.9, Measured Power Distribution Uncertainty, of the CASMO-4/MICROBURN-B2 codes system (Reference 4) provides comparison of the power distribution uncertainties between the upgraded CASMO-4/MICROBURN-B2 and the CASMO-3G/MICROBURN-B. In its September 16, 2003, letter (Reference 9), PPL notified the NRC of their plan to use the CASMO-4/MICROBURN-B2 code system to perform the reload core and bundle design, shutdown margin analysis, and the SLMCPR input generation. PPL submitted this letter in accordance with the GL 83-11, Supplement 1, License Qualification for Performing Safety Analysis, prior to using an approved code or method to perform safety-related evaluations.

3.2 TS 2.1.1 Reactor Core Safety Limits (SLs)

PPL proposed to change the SLMCPR values in TS 2.1.1.2 for SSES-1 Cycle 14 U1C14 operation from 1.12 to 1.08 for two recirculation loop operation (TLO) and from 1.13 to 1.10 for single recirculation loop operation (SLO) with the reactor steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10 million pound-mass per hour (lbm/hr). SLMCPR is not considered as a safety concern for reactor operation below 785 psig dome pressure at core flow of less than 10 lbm/hr.

In its July 1, 2003, submittal (Reference 1), PPL states that the decrease in SLMCPR values from SSES-1 Cycle 13 (U1C13) to U1C14 occurs for the following reasons:

1.

Removal of an excess conservatism that is not required by NRC-approved methodology, and 2.

Incorporation of smaller power distribution uncertainties in the SLMCPR analysis that is based on NRC-approved CASMO-4/MICROBURN-B2 methodology.

3.2.1 Elimination of Safety Factor of 2 The previous NRC-approved computer code ANFB critical power correlation had a mean bias (in the predicted to measured critical power ratio) that was slightly greater than 1.0. With a predicted to measured ratio of 1.003, the ANFB correlation on the average would result in a predicted critical power that is slightly higher than the measured critical power. To compensate for the slight overprediction of the critical power, NRC staff approved the ANFB critical power correlation with a safety factor of 2 to be applied to the number of pins in boiling transition in calculating the SLMCPR. For the U1C14 SLMCPR calculation, PPL applies the computer code ANFB-10 critical power correlation which has a mean bias less than 1.0 (0.9985). Since the mean bias in the ANFB-10 correlation is conservative, the NRC staff approved the use of the ANFB-10 critical power correlation without applying the factor of 2 to the number of pins in boiling transition. For U1C13, PPL conservatively applied the factor of 2 to the ANFB-10 critical power correlation methodology. For U1C14, the SLMCPR was calculated without applying the factor of 2 to the number of pins that are in boiling transition. Therefore, PPL gains some margin in the SLMCPR calculation, using the the NRC-approved ANFB-10 correlation, while maintaining the requirement that 99.9% of the fuel bundles do not experience boiling transition.

3.2.2 Reduction in the Power Distribution Uncertainties For U1C13, PPL used the POWERPLEX-II code system. Therefore, the power distribution uncertainties included in the SLMCPR calculation for U1C13 correspond to the CASMO-3/MICROBURN-B power distribution uncertainties. For U1C14, PPL proposes to use the POWERPLEX-III core monitoring system which utilizes the CASMO-4/MICROBURN-B2 code system (EMF-2158(p)(A), revision 0). Due to advanced features implemented in MICROBURN-B2 (core simulator code), the calculated nodal and pin power distributions are more accurate relative to the earlier version of CASMO-3/MICROBURN-B. PPL states the CASMO4/MICROBURN-B2 core simulator implemented in the POWERPLEX-III core monitoring system and the associated radial and local power distribution uncertainties decrease for U1C14. The plant, fuel, and critical power ratio correlation uncertainties are incorporated into the NRC-approved Framatome-ANP SLMCPR calculations method, and the lower uncertainties contribute to the lower SLMCPR value for U1C14 as compared to U1C13.

The NRC staff determined that for U1C14, PPL used CASMO-3G to generate the lattice neutronic data. In addition, MICROBURN-B was also used to perform the steady state cycle simulation and to establish the base reactor condition and the corresponding bundle power distributions for a given burnup and rod pattern. The core monitoring system would employ the upgraded CASMO-4/MICROBURN-B2 code system. Moreover, in performing the SLMCPR calculation, Framatome used the upgraded code systems lower power distribution uncertainties in perturbing the key parameters from the base case for a given rod pattern and burnup state.

The smaller perturbation of the key power distribution parameters yields the reduction in the SLMCPR limit for U1C14. The NRC staff found that the use of different code systems results in inconsistencies in the power distribution uncertainties and could potentially lead to the calculation of nonconservative SLMCPR values for the cycle. The NRC staff requested PPL to demonstrate why the use of the core monitoring systems lower power distribution uncertainties, instead of the uncertainties corresponding to the code system used to generate the actual bundle pin radial and axial power distribution, yield a conservative SLMCPR value.

In a February 26, 2004, meeting (ADAMS accession nos. ML040630331 and ML040640525),

PPL addressed the NRC staffs above concerns. PPL demonstrated that for a given SSES-1 core design, rod pattern, and burnup condition, MICROBURN-B2 predicted a flatter radial power distribution than MICROBURN-B, despite the code systems lower radial power distribution uncertainty. In addition, PPL stated that since the high powered bundles that contribute to the SLMCPR are assumed to be operating at the SLMCPR value initially selected and at the bundle linear heat generation rate limit, the differences in the radial power distribution between the codes have a much lower influence on the SLMCPR calculation.

Therefore, the PPL presentation to the NRC staff demonstrated that for U1C14, the power distribution uncertainty associated with those generated with CASMO-3G/MICROBURN-B, although slightly larger than those uncertainties associated with CASMO-4/MICROBURN-B2, had very small effect on the overall U1C14 safety limit. That is, the SLMCPR calculation, which is in general deterministic with some statistical component in the process, was found to be insensitive to minor variations arising from small differences that are due to computer code changes. The end result of determining the number of rods contributing to the boiling transition remained the same.

Approval of PPLs use of different code systems is limited to the upcoming U1C14, in which PPL demonstrated in the February 23rd meeting that the calculated SLMCPR limit is conservative. For future cycles, PPL has stated that it would transition to using consistent code systems to generate the lattice spectrum depletion calculations and for simulating the core steady state conditions; and in the event of upgrading the code system, PPL will submit the appropriate request prior to implementation of the new methodology.

In its November 17, 2003, submittal (Reference 2), PPL provided a revised core composition for U1C14. These changes were necessary to address design changes related to control cell friction mitigation. Four twice-burned ATRIUM-10 fuel assemblies were discharged and replaced with 4 fresh fuel assemblies in order to maintain full power energy targets as a result of the rod pattern adjustments needed to address the control cell friction issues. PPL states the resulting SLO and TLO SLMCPR values remain unchanged from the values reported in their July 1, 2003, submittal and that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or AOOs. The NRC staff notes that this item is controlled by PPLs Core Operating Limits Report and does not require a TS amendment.

In its December 22, 2003, submittal (Reference 3), PPL states that based on previous safety limit evaluations performed by Framatome for SSES, the reduction in the U1C14 SLMCPR values are attributed to the following:

1.

Deleting the factor of 2 based on ANFB-10 correlation results in a decrease of -0.01 to

-0.02 for both TLO and SLO.

2.

The reduction in the power distribution uncertainties yields approximately a -0.02 to

-0.03 reduction in the SLMCPR for both TLO and SLO.

3.

Cycle-to-cycle variability contributes +0.01 to -0.01 for both TLO and SLO.

In addition, PPL states the U1C14 reload and SLMCPR analyses for TLO and SLO was performed within the applicability range of the ANFB-10 correlation, including the additional uncertainty for local peaking greater than 1.5 as specified in the NRC-approved ANFB-10 correlation safety evaluation, dated July 17, 1998. PPL states that the fuel will be operated within the ANFB-10 correlation range of applicability during U1C14 operation.

The NRC staff has evaluated PPLs submittals (References 1, 2, and 3) to determine whether the proposed changes to the SLMCPR values are justified and are acceptable. Based on the results of the review, the NRC staff finds the U1C14 SLMCPR values acceptable. The proposed U1C14 SLMCPR values will ensure that 99.9% of the fuel rods in the core will not experience boiling transition. The requirements of GDC-10 are met with respect to acceptable fuel design limits. The NRC staff also concludes that the justification for analyzing and detemining the SLMCPR value of 1.08 for TLO and 1.10 for SLO is acceptable, because PPL used appropriate cycle-specific parameters and NRC-approved licensing methodologies, analytical methods, and codes. Therefore, the NRC staff finds the proposed changes in TS 2.1.1.2 acceptable.

3.3 TS 4.2.1 Fuel Assemblies PPL proposed a change to TS 4.2.1 to indicate the use of a small amount of depleted uranium (tails) in the fuel rods, in addition to natural and slightly enriched uranium dioxide (UO2).

Depleted uranuim is defined as uranium with an initial U235 weight percent that is less than naturally occurring uranium, because it is a byproduct of the uranium enrichment process.

The NRC staff has reviewed the proposed change and finds it acceptable since there is no change to the composition of the fuel pellets containing tails material, (i.e., UO2) except a slight decrease in the amount of U235. In addition, the use of depleted uranium (tails) in the fuel rods does not affect the mechanical performance of the fuel rods; therefore, the NRC staff finds the proposed change to TS 4.2.1 acceptable.

3.4 TS 5.6.5 Core Operating Limits Report (COLR)

PPL proposed to delete TS 5.6.5.b.5 and TS 5.6.5.b.7 which are no longer required for the U1C14 analysis. In addition, PPL proposed to add an additional document EMF-2158(P)(A),

Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," to the list of approved analytical methods in TS 5.6.5.b. which enables its use in core physics analyses for U1C14 and future reloads.

The NRC staff has reviewed the justification for the proposed revision and the response to the NRC staffs request for additional information. Based on results of the review and the information provided in the February 26, 2004, meeting, the NRC staff concludes that the proposed TS revision is acceptable since the references that are removed pertain to the analysis methodologies used to analyze FRA-ANP 9x9-2 fuel which is no longer used in SSES-1. Furthermore, the addition of the reference document which contains FRA-ANP analysis methodology to TS 5.6.5.b is acceptable since it has already been approved by the NRC for the uses requested by PPL (namely, to support calculations of the cycle-specific parameters specified in TS 5.6.5). The NRC staff finds the proposed changes are in compliance with the guidance specified in GL 88-16.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (68 FR 46245). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

B.L. Shriver, PPL, letter to U.S. NRC, Susquehanna Steam Electric Station Proposed Amendment No. 256 to Unit 1 License NPF-14: MCPR Safety Limits and Reference Changes PLA-5638, July 1, 2003.

2.

B.L. Shriver, PPL, letter to U.S. NRC, Susquehanna Steam Electric Station Proposed Amendment No. 256 to Unit 1 License NPF-14: MCPR Safety Limits and Reference Changes Revised Core Composition Table PLA-5690, November 17, 2003.

3.

B.L. Shriver, PPL, letter to U.S. NRC, Susquehanna Steam Electric Station Request for Additional Information Regarding Proposed Amendment No. 256 to Unit 1 License NPF-14: MCPR Safety Limits and Reference Changes PLA-5702, December 22, 2003.

4.

EMF-2158 (P)(A), Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, October 1999.

5.

EMF-1997(P)(A), ANFB-10 Critical Power Correlation, July 1998.

6.

A.C. Thadani, NRC, letter to R.A. Copeland, Framatome, Acceptance for Referencing of Topical Report XN-NF-80-19(P)(A), Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology, August 1990.

7.

ANF-524(P)(A), Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors, November 1990.

8.

PL-NF-90-001, Supplement 2-A, Application of Reactor Analysis Methods for BWR Design and Analysis, May 1996.

9.

B.L. Shriver, PPL, letter to U.S. NRC, Susquehanna Steam Electric Station Notification of the Use of NRC Previously Approved Methodology in Accordance with Generic Letter 83-11 Supplement 1, PLA-5664, September 16, 2003.

Principal Contributor: Z. Abdullahi Date: March 9, 2004