RNP-RA/10-0096, Request for Technical Specifications Change Regarding Use of M5 Alloy Fuel

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Request for Technical Specifications Change Regarding Use of M5 Alloy Fuel
ML102990125
Person / Time
Site: Robinson Duke Energy icon.png
Issue date: 10/20/2010
From: Bernie White
Progress Energy Co, Progress Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/10-0096
Download: ML102990125 (19)
v • d • e


Text

Progress Energy 10 CFR 50.90 Serial: RNP-RA/I 0-0096 OCT 2 0 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING USE OF M5 ALLOY FUEL Ladies and Gentlemen:

In accordance with the provisions of the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power and Light Company, also known as Progress Energy Carolinas (PEC), Inc., is submitting a request for an amendment to the Technical Specifications (TS) contained in Appendix A of the Operating License for the H. B. Robinson Steam Electric Plant (HBRSEP),

Unit No. 2. The proposed amendment would modify Technical Specifications to permit the use of the AREVA alloy designated as M5 in fuel rod cladding and fuel assembly structural components in future operating cycles.

Specifically, changes are requested in the description of fuel assemblies specified in TS 4.2.1.

Additionally, changes are requested to the analytical methods referenced in TS 5.6.5.b. The changes to TS 5.6.5.b includes the addition of AREVA Reports BAW-10240(P)(A),

"Incorporation of M5 Properties in Framatome ANP Approved Methods," and EMF-2328, "PWR Small Break LOCA Evaluation Model S-RELAP5 Based," and the deletion of nine analytical methods that were previously approved but are no longer planned to be used, and therefore have not been analyzed for acceptability for M5 alloy fuel.

Attachment I provides an Affirmation pursuant to 10 CFR 50.30(b).

Attachment II provides a description of the current condition, a description and justification of the proposed change, a No Significant Hazards Consideration Determination, and an Environmental Impact Consideration.

Attachment III provides a markup of the current TS pages.

Attachment IV provides retyped pages for the proposed TS.

Progress Energy Carolinas, Inc.

Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

United States Nuclear Regulatory Commission Serial: RNP-RA/10-0096 Page 2 of 2 In accordance with 10 CFR 50.91(b), PEC is providing the State of South Carolina with a copy of the proposed license amendment.

There are no plant specific analyses included with this submittal. Plant specific analyses for M5 fuel are expected to be completed by February 2011 for the Realistic Large Break LOCA and by May 2011 for the Small Break LOCA.

PEC requests approval of the proposed license amendment by October 1, 2011, to allow for loading of M5 cladding fuel assemblies into the core during Refueling Outage 27, currently scheduled to begin on October 29, 2011. Implementation would occur following shutdown from Operating Cycle 27, but prior to loading M5 fuel cladding assemblies for Operating Cycle 28.

An exemption from the requirements of 10 CFR Part 50.46 and 10 CFR Part 50, Appendix K, which is also required to allow the use of M5 cladding, is being submitted via a separate letter.

If you have any questions concerning this matter please contact Mr. Curt Castell at (843) 857-1626.

Sincerely, B. C. White Manager - Support Services - Nuclear Attachments I. Affirmation II. Request for Technical Specifications Change Regarding the use of M5 Alloy Fuel III. Markup of Technical Specifications Pages IV. Retyped Technical Specifications Pages c: Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste management Section (SC)

Mr. A. Gantt, Chief, Bureau of Radiological Health (SC)

Mr. L. A. Rayes, NRC, Region II Mr. T. Orf, NRC Project Manager, NRR NRC Resident Inspectors, HBRSEP Attorney General (SC)

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/10-0096 Page 1 of 1 AFFIRMATION The information contained in letter RNP-RA/10-0096 is true and correct to the best of my information, knowledge, and belief; and the sources of my information are officers, employees, contractors, and agents of Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. I declare under penalty of perjury that the foregoing is true and correct.

Executed On: LOLZt{(0 M enjamin C. White Maayr- Support Services - Nuclear

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 1 of 6 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING USE OF M5 ALLOY FUEL Description of Current Condition Technical Specification (TS) 4.2.1, Fuel Assemblies, currently states, "The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material."

TS 5.6.5.b, Core Operating Limits Report (COLR), contains the listing of analytical methods that have been approved for use at HBRSEP, Unit No. 2, to determine core operating limits.

Included in this listing are the following methods that are no longer required:

  • 5.6.5.b.1 Core Design Methodology XN-75-27 (A)

" 5.6.5.b.4 Steamline Break Methodologies ANF-84-093(P)(A) and EMF-84-093(P)(A)

" 5.6.5.b.6 Small Break LOCA Methodology XN-NF-82-49(A)

" 5.6.5.b.7 Large Break LOCA Methodology EMF-2087 (P)(A)

" 5.6.5.b. 10 Critical Heat Flux Correlation Methodology ANF-1224 (A)

" 5.6.5.b.12 Westinghouse SBLOCA Methodology WCAP-10080-A

" 5.6.5.b.13 Westinghouse SBLOCA Methodology WCAP-10081-A

" 5.6.5.b.14 Westinghouse LOCA Transient Methodologies WCAP-8305 and WCAP-8301

" 5.6.5.b.15 Reference to NRC SER for Cycle 10 License Amendment 87 Additionally, TS 5.6.5.b does not currently contain two analytical methodologies that will be required to evaluate the use of M5 alloy fuel at HBRSEP, Unit No. 2.

Description and Justification of the Proposed Change The proposed change to TS 4.2.1, Fuel Assemblies, is to add M5 to the description of the fuel rod cladding material thereby permitting the use of zircaloy-4 or M5 cladding for future fuel cycles.

The first proposed change to TS 5.6.5, Core Operating Limits Report (COLR), is to delete the nine previously approved methodologies listed above. The line items will be replaced with the word "Deleted" to maintain the current numbering format.

The second proposed change to TS 5.6.5 is to add the following two NRC approved methodologies to allow for their use at HBRSEP, Unit No. 2:

  • BAW-10240 (P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 2 of 6

  • EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,"

approved version as specified in the COLR.

The current fuel cladding used at HBRSEP, Unit No. 2, is zircaloy-4, as stated in TS 4.2.1.

AREVA developed the M5 advanced fuel rod cladding and fuel assembly structural material.

M5 has demonstrated superior corrosion resistance and reduced creep relative to both standard and low-tin zircaloy-4. The properties of the M5 material relevant to fuel design and safety analysis are documented in BAW- 10227 (P)(A), Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel." This topical report was reviewed and approved by the NRC. The proposed revision adds M5 as an option to zircaloy-4 in TS 4.2.1 to allow for the use of this improved fuel cladding material in future cycles.

Technical Specifications 5.6.5.b lists the methodologies used in the design and safety analysis of core reloads. The Core Operating Limit Report submitted to the NRC for each cycle identifies those methods and revisions used in the effective COLR. AREVA assessed the impact of M5 cladding on AREVA methodologies used for core reload design in BAW- 10240(P)(A),

"Incorporation of M5 Properties in Framatome ANP Approved Methods." The NRC approved BAW-10240(P)(A) for use in license applications in a letter dated May 5, 2004.

BAW- 10240(P)(A) describes the incorporation of the NRC-approved M5 material properties into a set of NRC-approved AREVA mechanical analysis, small break loss-of coolant accident (SBLOCA), and non-loss-of-coolant accident (non-LOCA) methodologies. The large break LOCA methodology EMF-2103 includes the application of M5 cladding.

EMF-2328(P)(A) allows the use of the S-RELAP5 computer code for analysis of small break LOCA. It was approved by NRC Safety Evaluation Report in a letter dated March 15, 2001.

Changes to this methodology for analysis of fuel with M5 cladding are addressed in BAW-10240(P)(A).

The list of COLR methodologies in TS 5.6.5.b has been reviewed to identify those methodologies that are not specifically addressed by BAW-10240(P)(A) and must be retained during the transition from zircaloy-4 to M5. These methodologies are:

" XN-NF-621(A), "XNB Critical HeatFlux Correlation" (TS 5.6.5.b Item 9)

" XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events" (TS 5.6.5.b Item 2)

" ANF-89-151 (A), "ANF-RELAP Methodology for Pressurized Water Reactors:

Analysis of Non-LOCA Chapter 15 Events" (TS 5.6.5.b Item 18)

ANF-89-151 and XN-NF-84-73 (Non-LOCA)

BAW- 10227 (P)(A), Section 4, demonstrates that non-LOCA related topical reports do not require revision to address the use of M5 cladding since there is a negligible impact of the M5 cladding on non-LOCA transients. Therefore continued reference to these topical reports in the Technical Specifications is acceptable. EMF-23 10 is directly discussed in BAW- 10240 (P)(A) with the same conclusions as to the impact of M5 cladding.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 3 of 6 XN-NF-621 (Critical Heat Flux Correlation)

BAW-10240 (P)(A), Section 6.2.3, demonstrates that departure from nucleate boiling (DNB) or critical heat flux (CHF) related topical reports do not require revision to address the use of M5 cladding since there is a negligible impact of the M5 cladding on DNB. Therefore continued reference to this topical report in the HBRSEP Technical Specifications is acceptable.

Additionally, the Technical Specifications are non-specific in regard to fuel cladding composition for fuel stored in the spent fuel pool. However, it has been determined that the licensing basis spent fuel pool criticality analyses are bounding for both zircaloy-4 and M5 clad fuel.

Safety limits for future operating cycles at the HBRSEP will continue to be analyzed using methodologies approved by the NRC. Therefore, this proposed change will have no adverse impact on plant safety.

The NRC has approved similar license amendments for Ft. Calhoun Station (Docket No. 50 -

285), Palisades Nuclear Power Plant (Docket No. 50-255), Arkansas Nuclear One (ANO) Unit 1, Docket No. 50-313, and Sequoyah Nuclear Station (Docket No. 50-327).

No Significant Hazards Consideration Determination Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), is proposing changes to Appendix A, Technical Specifications (TS), of Facility Operating License No. DPR-23, for the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The proposed changes will revise the description of fuel assemblies specified in TS 4.2.1. Additionally, changes are requested to the analytical methods referenced in TS 5.6.5.b. The changes to TS 5.6.5.b includes the addition of AREVA Reports BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," and EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based," and the deletion of nine analytical methods that were previously approved but are no longer planned to be used, and therefore have not been analyzed for acceptability for M5 alloy fuel.

An evaluation of the proposed changes has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations, using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed license amendment adds a Nuclear Regulatory Commission approved analytical method, BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," used to determine the core operating limits, to Technical Specification (TS) 5.6.5.b and changes the description of fuel assemblies specified in TS 4.2.1 to allow use

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 4 of 6 of the M5 alloy. The proposed amendment does not affect the acceptance criteria for any Final Safety Analysis Report (FSAR) safety analysis analyzed accidents or anticipated operational occurrences. The proposed amendment does not involve operation of the required structures, systems or components (SSCs) in a manner different from those previously recognized or evaluated. As such, the proposed amendment does not increase the probability or consequences of an accident.

In addition, the proposed license amendment adds NRC approved methodology EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based." This change, by itself, does not impact the current design bases. The proposed change enables the use of new methodologies to re-analyze small break loss-of-coolant accidents. Revised analyses may either result in continued conformance within design bases, or may change the design bases. If design bases changes result from a revised analysis, then the specific design changes will be evaluated in accordance with HBRSEP, Unit No. 2, design change procedures and 10 CFR 50.59. Further, this part of the change does not involve physical changes to any plant structure, system, or component.

In addition, the proposed license amendment deletes nine analytical methods that were previously approved and listed in Section 5.6.5.b, but are no longer planned to be used. This change is administrative in nature as it removes methodologies that have become obsolete and hence have not been analyzed for acceptability with M5 fuel.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated.

Use of M5 fuel will not result in changes in the operation or configuration of the facility.

Topical reports BAW-10227(P)(A) and BAW-10240(P)(A) evaluate the material properties of the M5 alloy and conclude that they are similar or better than those of zircaloy-4.

Therefore, M5 fuel rod cladding will perform similarly to those fabricated from zircaloy-4, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident. No new failure mechanisms will be introduced by the changes being requested.

The proposed addition of EMF-2328(P)(A) does not involve any physical alteration of plant systems, structures, or components, other than allowing for fuel design in accordance with NRC-approved methodologies. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no change to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result, no new failure modes are being introduced by introduction of this methodology.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 5 of 6 Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the M5 alloy are not significantly different from those of zircaloy-4. M5 alloy is expected to perform similarly or better than zircaloy-4 for all normal operating and accident scenarios, including both loss-of-coolant accident (LOCA) and non-LOCA scenarios. The proposed changes do not affect the acceptance criteria for any FSAR safety analysis analyzed accidents or anticipated operational occurrences. All required safety limits would continue to be analyzed using methodologies approved by the Nuclear Regulatory Commission.

There is no impact on any margin of safety resulting from the incorporation of these new topical reports into the Technical Specifications. If design basis changes result from a revised analysis that uses these new methodologies, the specific design changes will be evaluated in accordance with HBRSEP, Unit No. 2, design change procedures and 10 CFR 50.59. Any potential reduction in the margin of safety would be evaluated for that specific design change.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

Based on the evaluation above, PEC concludes that the proposed change does not involve a significant hazards consideration.

Environmental Impact Consideration 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions for categorical exclusion from performing an environmental assessment. A proposed change for an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed change would not (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; (3) result in a significant increase in individual or cumulative occupational radiation exposure. Carolina Power and Light Company, also known as Progress Energy Carolinas (PEC), Inc., has reviewed this request and determined the proposed change 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 needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/10-0096 Page 6 of 6 Proposed Change Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), is proposing changes to Appendix A, Technical Specifications (TS), of Facility Operating License No. DPR-23, for the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. The proposed changes will revise the description of fuel assemblies specified in TS 4.2.1. Additionally, changes are requested to the analytical methods referenced in TS 5.6.5.b. The changes to TS 5.6.5.b includes the addition of AREVA Reports BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," and EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based," and the deletion of nine analytical methods that were previously approved but are no longer planned to be used, and therefore have not been analyzed for acceptability for M5 alloy fuel.

Basis The proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not involve a significant hazards consideration.
2. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated and does not result in the possibility of a new or different kind of accident. Based on the material composition of M5 and the operational experience from plants that have been using M5 fuel, the proposed change should have no significant impact on plant effluents from normal operation. Therefore, the proposed change does not result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite.
3. Based on the material composition of M5 and the operational experience from plants that have been using M5 fuel, the proposed change should not result in a significant increase in individual or cumulative occupational radiation exposures.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/10-0096 5 pages including title page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING USE OF M5 ALLOY FUEL MARKUP OF TECHNICAL SPECIFICATIONS PAGES

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The H. B. Robinson Steam Electric Plant, Unit No. 2 is located on the southwest shore of Lake Robinson, in northwest Darlington County, South Carolina. The site location is approximately 25 miles NW of Florence, 35 miles NNE of Sumter, and 56 miles ENE of Columbia, South Carolina.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy-4 or M5 fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Rod Cluster Control (RCC) Assemblies The reactor core shall contain 45 full length RCC assemblies. The control material shall be silver-indium-cadmium, as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; (continued)

HBRSEP Unit No. 2 4.0-1 Amendment No. 1-76

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and
8. Boron Concentration limit for Specification 3.9.1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:
1. XN 75 27-(A), "EXXon ulor M rDoign Mothds for

-NeuFtron*

Prossurioed- Wa.I;ter Reactors," approvod vorsin a spcfiod I in the- G-. Deleted

2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.
3. XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. S-toam. Line Break Methodology as deffined-by: DQeleted ANl 84 093(P)(A), "Stfoea*Mln Br.oak Meth*de-oegy for PW\Rr,,"

approved- voreionR -asspocifiod in the; CO R.

EM1093(P)(A), "Steam Line Broak Mothodology for PWRc,"

app....ed .versio.n as6 specified in the C* R.*

5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow," approved version as specified in the COLR.
6. XN NF 82 49(A), "EXXOn NucleIPar Corporation Evaluation Model EXEM PWVR Small B-rea-k Model," approved vrina pcfe iMthe 0L-R7 Deleted
7. r-hEF 2087 (P)(A), "SE-IIPWR 98: ECS Evaluation Model for WAIR IBLOCA Applications," app-roed version as sp-ci,;,, in the GGL-RDeleted (continued)

HBRSEP Unit No. 2 5.0-25 Amendment No. 242

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.
9. XN-NF-621(A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.
10. ANF 1221 (A), "Departure from Nucleato BoiinOg Correlation for High Thermal PerFormance Fuel," appDeVd verSion ae spocified in the GGItR.Deleted
11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.
12. WC.ARp10080 .A, "NOTRUMP, A Nodal Transient Small Break and G-.enal Netw-rk Cede," appr*e.d ,4r.... a specified in the GO.I=R 7Deleted
13. WCtAP 10081 A, "Westinghouse Small Broak EGCS Evaluation Model Using the NOT+RUMP code," approved version as specified in the
G-=R.Deleted 7
14. VWCAP 8301 (PrVprietaV,) and WCGAP 8305 (Nonp1oprietarl'),

"LOCTA IV PrFogram-: Loss of CoolaRt Tr*aFnient Analysis," app.rved vrion a3. specified in tho COLR.Deleted

15. "Sat*~ Evaluation by the Oftie of Nucloar ReactorF Regulation Related to Amndment No. 87 to Facility Operating License No. DPR 23, Caoelina Pewer & Light Co., H. B. RobinSOn Steam Electric Plant, Unit Wo. 2,Docket No. 50 261," UJSNRC, Washington, DG 20555, 7 Nov.

8&4-Deleted

16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," approved version as specified in the COLR.
17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR.
18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-26 Amendment No. 244

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.
20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.
21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs,"

approved version as specified in the COLR.

23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.
24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.
25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.
26. BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.
27. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," approved version as specified in the COLR.
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

HBRSEP Unit No. 2 5.0-27 Amendment No. 2-4-9

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/10-0096 5 pages including title page H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING USE OF M5 ALLOY FUEL RETYPED TECHNICAL SPECIFICATIONS PAGES

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The H. B. Robinson Steam Electric Plant, Unit No. 2 is located on the southwest shore of Lake Robinson, in northwest Darlington County, South Carolina. The site location is approximately 25 miles NW of Florence, 35 miles NNE of Sumter, and 56 miles ENE of Columbia, South Carolina.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy-4 or M5 fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Rod Cluster Control (RCC) Assemblies The reactor core shall contain 45 full length RCC assemblies. The control material shall be silver-indium-cadmium, as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; (continued)

HBRSEP Unit No. 2 4.0-1 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. Axial Flux Difference (AFD) limits for Specification 3.2.3; and
8. Boron Concentration limit for Specification 3.9.1.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:
1. Deleted
2. XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," approved version as specified in the COLR.
3. XN-NF-82-21 (A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

approved version as specified in the COLR.

4. Deleted
5. XN-75-32(A), "Computational Procedure for Evaluating Rod Bow," approved version as specified in the COLR.
6. Deleted
7. Deleted
8. XN-NF-78-44(A), "Generic Control Rod Ejection Analysis," approved version as specified in the COLR.
9. XN-NF-621 (A), "XNB Critical Heat Flux Correlation," approved version as specified in the COLR.
10. Deleted
11. XN-NF-82-06(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup," approved version as specified in the COLR.
12. Deleted
13. Deleted (continued)

HBRSEP Unit No. 2 5.0-25 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

14. Deleted
15. Deleted
16. ANF-88-054(P), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," approved version as specified in the COLR.
17. ANF-88-133 (P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU," approved version as specified in the COLR.
18. ANF-89-151(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," approved version as specified in the COLR.
19. EMF-92-081 (A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.
20. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.
21. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

22. EMF-96-029(P)(A), "Reactor Analysis System for PWRs,"

approved version as specified in the COLR.

23. EMF-92-116, "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.
24. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.

(continued)

HBRSEP Unit No. 2 5.0-26 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

25. EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," approved version as specified in the COLR.
26. BAW-10240(P)(A), "Incorporation of M5 Properties in Framatome ANP Approved Methods," approved version as specified in the COLR.
27. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," approved version as specified in the COLR.
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(continued)

HBRSEP Unit No. 2 5.0-27 Amendment No.

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