ML090360738

Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies, and Proposed Change Number (PCN)-589, Amendment Application Numbers 254 and 240, Respectively

ML090360738
Person / Time
Site:	San Onofre
Issue date:	01/30/2009
From:	Short M Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090360738 (110)

Text

i~ISOUTHERN CALIFORNIA Michael P. Short EDISON aVice President An EDISON INTERNATIONAL Company January 30, 2009 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Docket Nos. 50-361 and 50-362 Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies, and Proposed Change Number (PCN)-589, Amendment Application Numbers 254 and 240, respectively for Units 2 and 3 Request to Revise Technical Specification 5.7.1.5, "Core Operating Limits Report (COLR)"

San Onofre Nuclear Generating Station, Units 2 and 3

References:

1.

Letter From NRC (M. Fields) to SCE (H. Ray) dated August 10, 1992," Acceptance of Topical Report SCE-9001, 'PWR Reactor Physics Methodology Using CASMO-3/SIMULATE-3' For Use at San Onofre Nuclear Generating Station, Units 1, 2, and 3!(TAC Nos.

M77846, M77843, and M77844)"

2.

Letter From NRC (S. Dembek) to SCE (H. Ray) dated June 2, 1999, "San Onofre Nuclear Generating Station, Units 2 and 3 - Evaluation of Reload Analysis Methodology Technology Transfer (TAC Nos.

MA4289 and MA4290)"

Dear Sir or Madam:

Pursuant to 10 CFR 50.12, "Specific Exemptions," Southern California Edison Company (SCE) is requesting a temporary exemption from the requirements of 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models." In addition to the temporary exemption, pursuant to 10 CFR 50.90, SCE is requesting an amendment to Technical Specification 5.7.1.5, "Core Operating Limits Report." The change will allow the use of the CASMO-4 methodology to perform nuclear design calculations.

SCE is developing a lead fuel assemblies (LFAs, also known as lead test assemblies or lead use assemblies) program with AREVA NP.

P.O. Box 128 San Clemente, CA 92674-0128 (949) 368-3780 Fax: (949) 368-3770

Document Control Desk January 30, 2009 Under this program, up to sixteen LFAs manufactured by AREVA NP may be inserted into the SONGS Unit 2 core or potentially into the SONGS Unit 3 core. Currently, eight AREVA LFAs are scheduled for installation in Unit 2 Cycle 16, with use for up to three operating cycles (Cycles 16, 17, and 18). Unlike current fuel assemblies, the AREVA LFAs will contain M5 alloy cladding material and Gadolinia (gadolinium oxide) burnable absorbers. As described below, SCE is requesting a temporary exemption to allow the use of M5 alloy cladding and a change to Technical Specification (TS) 5.7.1.5. Use of gadolinium oxide burnable absorbers is already authorized in TS 4.2.1.

Temporary Exemption Request The temporary exemption is required to allow up to sixteen LFAs manufactured by AREVA NP with M5 alloy cladding fuel rods to be inserted into the Unit 2 or Unit,3 reactor core, beginning with the upcoming Unit 2 refueling outage (Cycle 16).

The use of M5 alloy cladding LFAs allows SCE to evaluate cladding for future fuel assemblies in order to eliminate grid to rod fretting fuel failures. Since the requirements in 10 CFR 50.46 specifically, and 10 CFR 50, Appendix K implicitly, refer to the use of Zircaloy or ZIRLO cladding, a temporary exemption is required to use fuel rods clad with an advanced zirconium-based alloy that is not Zircaloy or ZIRLO. The temporary, exemption request and justification for the temporary exemption are described in detail in Enclosure 1 to this letter.

This temporary exemption is similar to the temporary exemption approved by the NRC for Calvert Cliffs (ADAMS Accession Number ML030640137), the temporary exemption for Palo Verde Unit 1 (ADAMS Accession Number ML082730003), and the exemption (ADAMS Accession Number ML011280063) and amendment (ADAMS Accession Number ML011300351) for Three Mile Island Unit 1.

A list of the regulatory commitments resulting from the exemption request is provided in.

Request to Amend Technical Specification 5.7.1.5 The upgrade from CASMO-3 to CASMO-4 is needed to model the AREVA LFAs, as they will contain Gadolinia burnable absorbers.

The SCE reactor core physics methodology using CASMO-3 and SIMULATE-3 has been previously approved by the NRC (References 1 and 2). The proposed amendment would add the CASMO-4 computer code to the list of analytical methods that may be used to determine the core operating limits contained in TS 5.7.1.5. Similar to CASMO-3, CASMO-4 would be used in conjunction with SIMULATE-3. The amendment request is documented in Enclosure 3.

Document Control Desk January 30, 2009 This request requires NRC review and approval of the enclosed Southern California Edison topical report, SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes."

SCE-0901 documents the applicability of the proposed CASMO-4 methodology to SONGS. This report also demonstrates SCE's proficiency to set up input decks, execute the programs, and properly interpret the results using CASMO-4.

SCE has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22 (c)(9). to this letter contains the Description and No Significant Hazards Analysis for the proposed amendment.

The proposed amendment is neither exigent nor emergency. SCE requests approval of this license amendment request (LAR) with an allowance of 60 days for implementation of the approved amendment.

Should you have any questions, or require additional information, please contact Ms. Linda Conklin at (949) 368-9443.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on //g,200T (Oate)

Sincerely,

Document Control Desk January 30, 2009

Enclosures:

Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies : Commitments : Description and No Significant Hazards Analysis for Proposed Change NPF-10/15-589 Technical Specification 5.7.1.5, "Core Operating Limits Report (COLR)" San Onofre Nuclear Generating Station Units 2 and 3 Attachment A: Existing Technical Specification Page, Unit 2 Attachment B: Existing Technical Specification Page, Unit 3 Attachment C: Proposed Technical Specification Page (Redline and Strikeout),

Unit 2 Attachment D: Proposed Technical Specification Page (Redline and Strikeout),

Unit 3 Attachment E: Proposed Technical Specification Page, Unit 2 Attachment F: Proposed Technical Specification Page, Unit 3 : SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design, Codes," January 2009 cc:

E. E. Collins, Regional Administrator, NRC Region IV N. Kalyanam, NRC Project Manager, San Onofre Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 and 3 S. Y. Hsu, California Department of Public Health, Radiologic Health Branch

ENCLOSURE 1 Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies

Introduction The San Onofre Nuclear Generating Station (SONGS) Unit 2 or Unit 3 core consists of 217 fuel assemblies. Each fuel assembly consists of 236 fuel rods. The rods are arranged in a square 16 x 16 array. The fuel rods consist of slightly enriched uranium dioxide cylindrical ceramic pellets, encapsulated within a cylindrical Zircaloy or ZIRLO tube.

Part 50.46(a)(1)(i) of Title 10 of the Code of Federal Regulations [10 CFR 50.46(a)(1)(i)]

states in part:

"Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant, accidents are calculated."

10 CFR 50.46 continues on to delineate specifications for peak cladding temperature, maximum hydrogen generation, coolable geometry, and long-term cooling. Since 10 CFR 50.46 specifically refers to fuel with Zircaloy or ZIRLO cladding, the use of fuel clad with zirconium-based alloys that do not conform to either of these two designations requires a temporary exemption from this section of the regulations.

10 CFR 50, Appendix K, paragraph I.A.5, states:

"The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation."

The Baker-Just equation presumes the use of Zircaloy or ZIRLO cladding. The use of fuel with zirconium-based alloys that do not conform to either of these two designations requires a temporary exemption from this section of the regulations.

Pursuant to 10 CFR 50.12, "Specific Exemptions," Southern California Edison Company (SCE) is requesting a temporary exemption from the requirements of 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" and 10 CFR 50, Appendix K "ECCS Evaluation Models" for San Onofre Units 2 and 3.

The temporary exemption will allow a limited number (not to exceed 16) of lead fuel assemblies (LFAs, also known as lead test assemblies or lead use assemblies) manufactured by AREVA NP with M5 alloy clad fuel rods containing Gadolinia burnable Page 1 of 7

absorbers to be inserted into SONGS Unit 2 or 3 reactor core in nonlimiting core regions during the upcoming Cycle 16 refueling outage. The LFAs will remain in the core for up to three operating cycles. The use of M5 alloy cladding LFAs allows SCE to evaluate cladding for future fuel assemblies in order to eliminate grid to rod fretting fuel failures.

The regulations specify standards and acceptance criteria only for fuel rods clad with Zircaloy or ZIRLO. Thus a temporary exemption is required to use fuel rods clad with an advanced alloy that is not Zircaloy or ZIRLO.

Currently, at least ten US plants have used M5 alloy cladding fuel assemblies either in full batch or LFA programs, including Arkansas Nuclear One Unit 1, Crystal River Unit 3, Three Mile Island Unit 1, Davis Besse, North Anna Units 1 and 2, Sequoyah Units 1 and 2, Oconee Unit 2, Ft. Calhoun, Palo Verde Unit 1, and Braidwood Unit 1. M5 alloy cladding fuel assemblies have also been used extensively in European plants.

10 CFR 50.12, Specific Exemption The standards set forth in 10 CFR 50.12 provide that the Commission may grant exemptions from the requirements of the regulations of this part for reasons consistent with the following:

The exemption is authorized by law; The exemption will not present an undue risk to the public health and safety;

" The exemption is consistent with the common defense and security; and

" Special circumstances are present.

This exemption is authorized by law. The remaining standards for the temporary exemption are also satisfied, as described in the following paragraphs.

The exemption will not present an undue risk to public health and safety. The NRC-approved M5 topical report (Reference 1) demonstrates that predicted chemical, mechanical, and material performance characteristics of the M5 alloy cladding are within those approved for Zircaloy under anticipated operational occurrences and postulated accidents. The LFAs will be placed in nonlimiting core regions as required by Technical Specification (TS) 4.2.1 "Fuel Assemblies." In the unlikely event that cladding failures occur in the LFAs, the environmental impact would be minimal and is bounded by previous accident analyses. Therefore, the use of the advanced zirconium-based cladding material, M5 alloy, will not present an undue risk to the public health and safety.

The exemption is consistent with the common defense and security. The use of M5 alloy clad LFAs allows SCE to evaluate this cladding material for use in future fuel assemblies, to provide a more robust design to eliminate grid to rod fretting fuel failures.

Special circumstances are present. As set forth in 10 CFR 50.12(a)(2)(ii), which states that special circumstances are present whenever "Application of the regulation in Page 2 of 7

the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule...".

10 CFR 50.46 identifies acceptance criteria for ECCS system performance at nuclear power facilities. The effectiveness of the ECCS in SONGS Units 2 and 3 will not be affected by the insertion of the LFAs. Due to the similarities in the material properties of the M5 alloy to Zircaloy or ZIRLO as identified in the AREVA M5 alloy topical report (Reference 1) and the placement of the LFAs in nonlimiting core regions, it can be concluded that the ECCS performance would not be adversely affected.

The intent of paragraph I.A.5 of Appendix K to 10 CFR 50 is to apply an equation for rates of energy release, hydrogen generation, and cladding oxidation from metal-water reaction that conservatively bounds all post-LOCA (Loss of Coolant Accident) scenarios. The supporting documentation for the AREVA M5 topical (Reference 1) shows that due to the similarities in the composition of the M5 alloy cladding and Zircaloy or ZIRLO, the application of the Baker-Just equation will continue to conservatively bound all post-LOCA scenarios.

A strict interpretation of 10 CFR 50.46 and 10 CFR 50, Appendix K, would not allow the use of M5 alloy clad fuel rods in lead fuel assemblies since the cladding material does not fall within the strict definition of Zircaloy or ZIRLO even though the AREVA M5' topical report shows that the intent of the regulations is met. Application of these regulations in this particular circumstance would not serve the underlying purpose of the rule and is not necessary to achieve the underlying purpose of the rule, so special circumstances exist.

Lead Fuel Assembly (LFA) Program Summary SCE is requesting this temporary exemption in order to install LFAs manufactured by AREVA NP in the SONGS Unit 2 core or potentially into the SONGS Unit 3 core. The AREVA NP mechanical design for the SONGS LFAs is similar to the standard AREVA NP high thermal performance (HTP) fuel designed for Combustion Engineering (CE) 14x14 fuel pin lattice reload fuel, with the primary difference being the 16x16 fuel pin lattice. The M5 cladding will be used on the Calvert Cliffs 14x14 CE design, beginning in year 2010. The AREVA LFA design for SONGS is very similar to the AREVA LFAs being evaluated at Palo Verde (ADAMS Accession Number ML082730003). The mechanical design evaluations for the LFAs will be performed with the standard reload mechanical design methods using the M5 alloy cladding properties.

The NRC has reviewed and approved (Reference 2) the M5 alloy cladding properties in topical report BAW-10227P-A (Reference 1).

The LFAs are currently scheduled for installation in Cycle 16 of SONGS Unit 2, for up to three cycles of irradiation (Cycles 16, 17, and 18). The burnup achieved after three cycles of irradiation will be less than the current NRC approved (Reference 3) San Onofre burnup limit of 60 MWd/kgU, which is also less than the approved AREVA NP methodology peak rod limit of 62 MWd/kgU as described in report ANF-88-133(P)(A)

Page 3 of 7

(Reference 4). Prior to use of AREVA LFAs for a second or third fuel cycle of irradiation, poolside LFA examinations will be performed to evaluate assembly and cladding performance, and acceptability for continued use (Commitments 1 and 2). If the AREVA LFAs are inserted for a third fuel cycle of irradiation, then poolside LFA examinations will be performed after completion of the third fuel cycle of irradiation to evaluate assembly and cladding performance (Commitment 3).

The fuel management will place the LFAs in nonlimiting core regions. Since these assemblies will not be in the highest power density locations, the placement scheme will assure that the behavior of the LFAs is bounded by the safety analyses performed for the standard fuel rods.

SCE, AREVA, and Westinghouse evaluations will verify performance of the LFAs with respect to the safety analysis. The analyses will include thermal hydraulic compatibility, LOCA and non-LOCA criteria, mechanical design, seismic and core physics. The evaluations will make use of the fact that the LFAs will be operated in nonlimiting core regions and will verify that the reload analyses are not adversely impacted. In addition, an evaluation will be performed to verify the insertion of the AREVA LFAs does not adversely impact the fuel performance and mechanical integrity of the co-resident Westinghouse fuel.

LFA Mechanical Design Description The LFAs for the SONGS Unit 2 or Unit 3 reactor will be the AREVA NP 16 x 16 CE design. The fuel bundle uses ten M5 spacer grids of the high thermal performance (HTP) design and one Alloy 718 spacer grid of the high mechanical performance (HMP) design. The lower end fitting is the FUELGUARD design, and the upper end fitting is a reconstitutable AREVA NP design for CE fuel. The HTP spacer grid was generically reviewed and accepted by the NRC and has been used in reload design for CE, Westinghouse, and Kraftwork Union reactors since 1991. The FUELGUARD lower end fitting has also been used in reload design for CE, Westinghouse, and General Electric reactors. The reconstitutable upper end fitting design has been used in reload design for plants with CE 14 x 14 fuel pin lattices.

Each fuel bundle contains four outer guide tubes, one center guide/instrument tube, and 236 fuel rods. The LFA fuel rods have the same pellet stack height and overall dimension as the co-resident Westinghouse fuel.

The primary differences between the Westinghouse fuel design currently used in the SONGS reactors and the AREVA fuel design include (1) the use of the different zirconium-based alloys for fuel rod cladding, guide tubes, and spacer grids; (2) the use of HTP grids; and (3) the use of a different burnable absorber (Gadolinia).

LFA in Nonlimiting Core Regions SONGS TS 4.2.1 "Fuel Assemblies," states:

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"The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include: borosilicate glass -

Na20-B 20 3-SiO2 components, boron carbide - B 4C, zirconium boride - ZrB2, gadolinium oxide - Gd203, erbium oxide - Er20 3. Limited substitutions of zirconium alloy (such as ZIRL OTM or Zircaloy) 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."

SCE is currently planning to place eight AREVA LFAs in nonlimiting core regions in Cycle 16. SCE has defined nonlimiting core regions as core locations where the peak integrated radial power peaking factor in the LFAs will be 0.95 or less of the core maximum integrated radial power peaking factor at all times in life (Commitment 4).

Therefore, the LFAs will not contain the limiting rod in the core and will have margin relative to the bounding peaking factors used in safety analyses. This criterion is consistent with that used in the AREVA LFA program at Palo Verde (ADAMS Accession Number ML082620212). The AREVA LFAs will be explicitly modeled in the SONGS core physics models and their impact will be analyzed in the cycle-specific core physics calculations that support the reload analyses (Commitment 5). SCE will design the LFA lattice, Cycle 16 core loading pattern, and perform reload physics analyses using both the Westinghouse (ABB-CE) reload methodology (Reference 5) and STUDSVIK CASMO and SIMULATE Code packages. The SONGS Cycle 16 core design will ensure that the LFA predicted peak rod power meets the 0.95 criterion.

An underlying assumption of the LFA program is that a 5% radial power peaking penalty will be sufficient to ensure that the LFAs will be nonlimiting in the safety, fuel performance, thermal-hydraulic, and Emergency Core Cooling System (ECCS) performance analyses. The 0.95 radial power peaking factor criterion applied to the LFA is a means of applying the 5% penalty to the LFAs. Since the LFAs will not be in the highest core power density locations, the placement scheme assures that the behavior of the LFAs is bounded by the safety analyses performed for the co-resident Westinghouse fuel. Additionally, the maximum LFA integrated fuel rod burnup will be maintained less than or equal to 60 MWd/kgU, the SONGS limit in Updated Final Safety Analysis Report (UFSAR) section 4.2.3.2.12.4 "SONGS Burnup Extension to 60,000 MWD/MTU."

For subsequent cycles containing the LFAs (Cycle 17 and potentially Cycle 18), SCE will explicitly model and analyze the AREVA LFAs in the reload core physics analyses, in the same manner as is performed for the lead cycle, Cycle 16. In addition to being once or twice burned, the reload core design for these cycles will determine the physical Page 5 of 7

peaking factors lower than the 0.95 criterion, thereby ensuring that the LFAs remain in nonlimiting core regions (Commitment 4).

AREVA LFA Analyses AREVA will perform detailed design analyses for the LFAs, including thermal-hydraulic compatibility, LOCA and non-LOCA criteria, mechanical design, thermal hydraulic, and seismic analyses of the AREVA LFAs in the SONGS reactor core (Commitment 6). The analyses will make use of the fact that the LFAs will be operated in nonlimiting core regions and will verify the reload analyses are not adversely impacted.

These analyses will include tasks such as physical design, normal and faulted operations, growth calculations, pressure drop and flow testing, among others. In addition, AREVA will analyze the seismic performance of the LFAs by evaluating the seismic/LOCA time history with respect to the strength of the AREVA LFAs.

The AREVA LFA analyses require the resources and coordination of SCE, AREVA, and Westinghouse. SCE will maintain overall responsibility for the LFA project. SCE has entered a three-party proprietary information agreement with AREVA and Westinghouse which allows the exchange of technical information for this project to ensure the-compatibility of the LFAs with the SONGS co-resident Westinghouse fuel and core internals. SCE acts as the intermediary between AREVA and Westinghouse to ensure that each organization has the necessary and appropriate inputs to perform the analyses.

Co-Resident Westinghouse Fuel Compatibility Analyses Westinghouse will perform a compatibility analysis to ensure that insertion of the AREVA LFAs will not cause the remaining Westinghouse fuel to exceed its operating limits and ensure there is no adverse impact on the fuel performance or mechanical integrity (Commitment 7).

Poolside LFA Examinations Poolside LFA examinations to assess key performance measures will include, as a minimum, 4-face inspections of the highest burn LFAs. Based on results of the inspection, the inspection scope may be expanded. Additional scope inspections could include but are not limited to additional visual inspections, oxide/crud lift-off measurements, fretting and diameter measurements, shoulder gap, assembly length and guide tube wear measurements.

Page 6 of 7

Precedent This temporary exemption request is similar to the temporary exemption approved by the NRC for Calvert Cliffs (ADAMS Accession Number ML030640137) and Palo Verde (ADAMS Accession Number ML082730003), and to the amendment and related exemption for Three Mile Island Unit 1 (ADAMS Accession Number ML011300351).

References

1. BAW-10227P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, Framatome Cogema Fuels, February 2000 (proprietary)

(ADAMS Ascension Number ML003686305).

2. NRC Revised Safety Evaluation (SE) for Topical Report BAW-1 0227P: "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" (TAC NO. M99903) dated February 4, 2000 (ADAMS Ascension Numbers ML003681479 and ML003681490).

3. NRC Generic Approval Of C-E Topical Report CEN-386-P, "Verification of the-Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x16 PWR Fuel" (TAC NO. M82192) dated June 22, 1992 (proprietary).

4. ANF-88-133(P)(A) and Supplement 1, Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62GWd/MTU, Advanced Nuclear Fuels Corporation, December 1991 (proprietary).

5. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3", June 1999 (proprietary).

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ENCLOSURE 2 Commitments

Commitments

1.

Prior to use of AREVA Lead Fuel Assemblies (LFAs) for a second fuel cycle of irradiation, poolside LFA examinations will be performed to evaluate assembly and cladding performance, and acceptability for continued use.

2.

Prior to use of AREVA LFAs for a third fuel cycle of irradiation, poolside LFA examinations will be performed to evaluate assembly and cladding performance, and acceptability for continued use.

3.

If the AREVA LFAs are inserted for a third fuel cycle of irradiation, then poolside LFA examinations will be performed after completion of the third fuel cycle of irradiation to evaluate assembly and cladding performance.

4.

The AREVA LFAs will be placed in core locations where the peak integrated radial power peaking factor in the LFAs will be 0.95 or less of the core maximum integrated radial power peaking factor at all times in life.

5.

The AREVA LFAs will be modeled in the SONGS core physics models and their impact will be analyzed in the cycle-specific core physics calculations that support the reload analyses.

6.

Analyses will be performed to verify the performance of the AREVA LFAs. These analyses include thermal-hydraulic compatibility, loss-of-coolant accident (LOCA) and non-LOCA criteria, mechanical design, thermal hydraulic, seismic, core physics, and neutronics compatibility of the AREVA LFAs in the SONGS reactor core. The analyses will make use of the fact that the LFAs will be operated in nonlimiting core regions and will verify the reload analyses are not adversely impacted.

7.

A compatibility analysis will be performed to ensure that insertion of the AREVA LFAs will not cause the remaining Westinghouse fuel to exceed its operating limits and ensure there is no adverse impact on the fuel performance or mechanical integrity.

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ENCLOSURE 3 Description and No Significant Hazards Analysis for Proposed Change NPF-10/15-589 Technical Specification 5.7.1.5, "Core Operating Limits Report (COLR)"

San Onofre Nuclear Generating Station Units 2 and 3

Description and No Significant Hazards Analysis for Proposed Change NPF-10/15-589 Technical Specification 5.7.1.5, "Core Operating Limits Report (COLR)"

San Onofre Nuclear Generating Station Units 2 and 3 EXISTING TECHNICAL SPECIFICATIONS Unit 2: see Attachment A Unit 3: see Attachment B PROPOSED TECHNICAL SPECIFICATIONS (highlight for additions, strikeout for deletions)

Unit 2: see Attachment C Unit 3: see Attachment D PROPOSED TECHNICAL SPECIFICATIONS (with changes)

Unit 2: see Attachment E Unit 3: see Attachment F

1.0 INTRODUCTION

Southern California Edison Company (SCE) requests NRC approval to incorporate use of the Studsvik-Scandpower computer program CASMO-4 (Reference 1) to support physics design analyses.

The currently approved SCE reload analysis methodology uses CASMO-3 (Reference 2) for cross-section generation and SIMULATE-3 (Reference 3) for core simulation. The current SCE reload analysis methodology is documented in SCE-9801-P-A (Reference 4) and SCE-9001-A (Reference 5).

The CASMO-3 and SIMULATE-3 methodologies have been used to evaluate reload designs at San Onofre for over fifteen years. SCE proposes to allow the use of CASMO-4 to support physics design analyses. CASMO-4 will be used in conjunction with SIMULATE-3. CASMO-4 streamlines the cross-section generation process while maintaining current peaking factor uncertainties. CASMO-4 was benchmarked against Page 1 of 7

CASMO-3 and it was determined that CASMO-4 does not reduce previously approved safety margins. SCE physics analysis methodology report SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes" (Reference 6), is enclosed for NRC review and approval.

2.0 PROPOSED CHANGE

SCE requests NRC approval to use the CASMO-4 computer program for San Onofre physics analyses to streamline the process of generating cross-sections and to reduce the potential for human error. SCE has prepared topical report SCE-0901 to support physics analyses using theICASMO-4 computer program.

TS 5.7.1.5, "Core Operating Limits Report (COLR)", will be revised by adding new Reference 9 to Section (b) as follows:

9.

SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

3.0 BACKGROUND

Studsvik has developed a new cross-section generation program called CASMO-4 since the initial development of SCE's reload analysis methodology. CASMO-4, like CASMO-3, is a two-dimensional neutron transport theory lattice physics code with depletion capability and the ability to generate cross-sections and discontinuity factors for boiling water reactor (BWR) and pressurized water reactor (PWR) diffusion theory physics analysis. SCE will use CASMO-4 for cross-section generation. The cross-sections from CASMO-4 will be used in SIMULATE-3 to generate physics parameters used to perform safety analyses.

CASMO-4 incorporates the microscopic depletion of burnable absorbers into the main calculations, which replaces the MICBURN-3 auxiliary code needed by CASMO-3. A new feature of CASMO-4 is the use of the characteristics form for solving the transport equation, which allows for a true heterogeneous geometry in the two-dimensional calculation. Studsvik has also automated the cross-section generation process with CASMO-4 so that complete nuclear data for SIMULATE-3 can be generated in one execution, which results in reduced potential for human error in the model development process.

CASMO-4 is in widespread use in the nuclear industry. Refer to Section 7.0 for examples of where the NRC has specifically approved the use of CASMO-4.

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4.0 TECHNICAL ANALYSIS

SCE has performed benchmarking of CASMO-4 against plant-specific measured data and against data calculated with CASMO-3 from Reference 5. Studsvik-Scandpower has performed benchmarking of CASMO-4 against critical experiments, including Gadolinia as the burnable absorber. The CASMO-4 benchmarking results are included in SCE-0901 (Reference 6).

The Critical Boron Concentrations, CEA worths, and Isothermal Temperature Coefficients generated with the CASMO-4 model show good general agreement with measurements. The peaking factors calculated with the CASMO-4 model also show good agreement with those calculated with the CASMO-3 model. Also, the peaking factor uncertainties are not changing as a result of CASMO-4. Thus, the newer version of CASMO does not reduce previously approved margins.

As discussed in SCE-0901, CASMO-4 can accurately model Gadolinia burnable absorbers. This capability is important to SCE since the AREVA lead fuel assemblies (LFAs) use gadolinia as the burnable absorber.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration SCE has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment", as discussed below.

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

SCE is adding the CASMO-4 computer program to its physics analysis methodology and will use the program for nuclear design analysis. This will allow the use of the CASMO-4 methodology to perform all steady-state pressurized water reactor (PWR) nuclear design analyses. The probability of occurrence of an accident previously evaluated will not be increased by the proposed change in the particular computer programs used for physics calculations for nuclear design analysis. The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the Updated Final Safety Analysis Report (UFSAR). These inputs do not alter physical characteristics or modes of operation of any system, structure, or Page 3 of 7

component involved in the initiation of an accident. Thus, there is no significant increase in the probability of an accident previously evaluated as a result of this change.

The consequences of an accident evaluated in the UFSAR are affected by the values of the physics inputs to the safety analysis. An extensive benchmark of CASMO-4 was performed with both San Onofre measured and predicted data, and with critical experiments. The accuracy of the CASMO-4 model is similar to the accuracy of the CASMO-3 model. Furthermore, there is the potential for the value of the nuclear design parameters to change solely as a result of the new core reload full core loading pattern. Regardless of the source of a change, an assessment is made of changes to the nuclear design parameters with respect to their effects on the consequences of accidents previously evaluated in the UFSAR. Thus, the nuclear design parameters are intermediate results and by themselves will not result in an increase in the consequences of an accident evaluated in the UFSAR.

Therefore, the use of the CASMO-4 methodology, which will perform the same functions as the existing CASMO-3 methodology with similar accuracy, does not significantly increase the consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The possibility for a new or different kind of accident evaluated previously in the UFSAR will not be created by the change to the particular methodologies used for physics calculations for nuclear design analyses. The change involves adding CASMO-4 to the SCE physics analysis methodology. CASMO-4 is an update to the CASMO-3 methodology currently approved for use at San Onofre.

The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the UFSAR. These inputs do not alter the physical characteristics or modes of operation of any system, structure or component involved in the initiation of an accident. Therefore, the addition of CASMO-4, which will perform the same functions as CASMO-3 with similar accuracy, does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 4 of 7

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response

No.

The proposed change does not involve a significant reduction in a margin of safety. The margin of safety as defined in the basis for any technical specification will not be reduced by the proposed change to the computer programs used for physics calculations for nuclear design analyses.

The change involves the addition of CASMO-4 to the SCE physics analysis methodology for nuclear design analysis. Extensive benchmarking of CASMO-4 has demonstrated that the values of those parameters used in the safety analysis are not significantly changed relative to the values obtained using the NRC approved CASMO-3 methodology. For any changes in the calculated values that do occur, the application of appropriate biases and uncertainties ensures that the current margin of safety is maintained. Specifically, use of these code specific biases and uncertainties in safety analyses continues to provide the same statistical assurance that the values of the nuclear parameters used in the safety analysis are conservative with respect to the actual values.on at least a 95/95 probability/confidence basis.

Based on the above, SCE concludes that the proposed amendment presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria SCE's NRC approved physics analysis methodologies, described in Reference 4 and Reference 5, meet applicable regulatory requirements including the following 10 CFR 50 Appendix A general design criteria:

Criterion 10- Reactor design. 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 during any condition of normal operation, including the effects of anticipated operational occurrences.

Criterion 11 - Reactor inherent protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

Criterion 12 - Suppression of reactor power oscillations. The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified Page 5 of 7

acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

Including the CASMO-4 computer program in SCE's physics analysis methodology does not change SCE's compliance with applicable regulatory requirements.

5.3 Conclusion SCE evaluated the CASMO-4 computer program and determined that calculations performed with CASMO-4 yield similar results to its predecessor CASMO-3.

CASMO-4 also does not change uncertainties associated with the peaking factors.

Thus, CASMO-4 does not reduce the previously approved margin of safety.

The change does not alter, degrade, or prevent actions described or assumed in any accident analysis in the UFSAR. It will not change any assumptions previously made in evaluating radiological consequences or affect any fission product barriers, nor does it increase or have any impact on the consequences of events described and evaluated in Chapter 15 of the UFSAR.

In conclusion, based on the considerations discussed above, (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 Commission's 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.

6.0 ENVIRONMENTAL EVALUATION Based on the above considerations, the proposed amendment does not involve and will not result in a condition which significantly alters the impact of San Onofre on the environment. Thus, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51.22(c)(9), and pursuant to 10 CFR Part 51.22(b),

no environmental assessment need be prepared.

7.0 PRECEDENT Examples of where the NRC has specifically approved the use of CASMO-4 include McGuire (ADAMS Accession Number ML082820015), Catawba (ADAMS Accession Number ML082820047), Palo Verde (ADAMS Accession Number ML010860187), and North Anna and Surry (ADAMS Accession Number ML030700038) and Fort Calhoun (ADAMS Accession Number ML050750534).

Page 6 of 7

8.0 REFERENCES

1.

SSP-01/400 Rev 4, Studsvik-Scandpower Inc, December 2004, "CASMO A Fuel Assembly Burnup Program, User's Manual"

2.

STUDSVIK/NFA-89/3, Studsvik Energiteknik AB, Sweden, November 1989, "CASMO A Fuel Assembly Burnup Program, User's Manual, Version 4.4"

3.

STUDSVIK/SOA-92/01, Studsvik of America Inc, April 1992, "SIMULATE Advanced Three-Dimensional Two-Group Reactor Analysis Code, Version 4.02"

4.

SCE-9801-P-A Rev 0, Southern California Edison, June 1999, "Reload Analysis Methodology For The San Onofre Nuclear Generating Station, Units 2 and 3"

5.

SCE-9001-A, Southern California Edison "PWR Reactor Physics Methodology Using CASMO-3/SIMULATE-3," September 1992

6.

SCE-0901, Southern California Edison "PWR Reactor Physics Methodology Using Studsvik Design Codes," January 2009 Page 7 of 7

Attachment A Existing Technical Specification Page, Unit 2

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16,

1986, G. W. Knighton (NRC)"'to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle:,.3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License NPF-1O and Amendment No.

19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

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 System 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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented-in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. 19-7,99,203 1

Attachment B Existing Technical Specification Page, Unit 3

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertai nti es"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16,

1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycl]e 3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License NPF-1O and Amendment No.

19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

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 System 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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No.

188,19n,195

Attachment C Proposed Technical Specification Page, Redline and Strikeout, Unit 2.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15," San Onof,,e, Nuclear Generating Station Units 2 and 3 (Cycle.3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License NPF-1O and Amendment No.

19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A L9.

SCE-0901, "PWýR Reactor Physics Methodology Using*

Studsvik Design Code~s" /

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 System 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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. 19*,19*-2,

Attachment D Proposed Technical Specification Page, Redline and Strikeout, Unit 3

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertai nti es"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15,"

San Onofre Nuclear Generating Station Units 2 and 3 (Cycl~e-3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License.NPF-1O and Amendment No.

19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOT Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9.

SCE-O901, "PWR Reactor Physics Methodology Using*

Studsvik Design Codes"

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 System 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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No. 1o8o8,19,19-

Attachment E Proposed Technical Specification Page, Unit 2

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluati-Pn Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15,"

San Onofre.

Nuclear Generating Station Units 2 and 3 (Cycle,.3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License NPF-1O and Amendment No.

19 to Facility Operating License NPF-15,"

San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD.404-P-A

9.

SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

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 System 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 mid-cycle revisions or supplements,,

shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No.

Attachment F Proposed Technical Specification Page, Unit 3

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S),

"Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16,

1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE),

"Issuance of Amendment No.

47 to Facility Operating License NPF-1O and Amendment No.

36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No.

30 to Facility Operating License NPF-1O and Amendment No.

19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8.

"Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9.

SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

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 System 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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No.

ENCLOSURE 4 SCE-0901, PWR Reactor Physics Methodology Using Studsvik Design Codes

SCE-0901 PWR REACTOR PHYSICS METHODOLOGY USING STUDSVIK DESIGN CODES January 2009 SOUTHERN CALIFORNIA EDISON COMPANY

SCE-0901 PWR REACTOR PHYSICS METHODOLOGY USING STUDSVIK DESIGN CODES C ") J. e. a Prepared By:

Date C. W. Gabel, Senior Engineer Reviewed By:

Approved By:

Approved By:

A. Bencheikh, Senior Engineer Superviqr Nluclear Fuel Management R. Y. Chang Supervisor, Nuclear Fuel Management (211( Ze (0 q Date Date Approved By:

7Da9 r, Nuclear Fuel Management

/

SCE-0901 DISCLAIMER This document was prepared by Southern California Edison Company for its own use. The use of information contained in this document by anyone other than Southern California Edison Company is not authorized, and in regard to unauthorized use neither Southern California Edison Company or any of its officers, directors, agents, or employees assumes any obligation, responsibility or liability, or makes any warranty or representation, with respect to the contents of this document, or its accuracy or completeness.

ii

SCE-0901 ABSTRACT This report documents the validation and level of accuracy of the reactor core physics methodology used by Southern California Edison Company to perform steady-state analyses for for Pressurized Water Reactors (PWR). In June 1992, SCE obtained NRC approval to use CASMO-3 / SIMULATE-3 (Reference 1) for reactor core physics design activities. SCE is upgrading from CASMO-3 (and supporting programs) to CASMO-4 (and supporting programs).

The CASMO-4 / SIMULATE-3 methodology has been validated by an in-house benchmarking effort against, predictions with measured data, critical experiments, a higher order computer program (MCNIP), and analyses of record (CASMO-3 / SIMILATE-3 results). Based on the results from this benchmarking effort, a set of 95/95 tolerance limits (uncertainties) has been calculated. Southern California Edison Company intends to use this methodology to perform PWR calculations including reload design, input to safety analyses, startup predictions, core physics databooks, and reactor protection system and monitoring system setpoint updates.

iii

SCE-0901 TABLE OF CONTENTS Page Disclaimer ii Abstract ii1 List of Tables vi List Of Figures viii List Of Acronyms ix SECTION 1 INTRODUCTION, OVERVIEW, AND

SUMMARY

1 1.0 Introduction 1.1 Overview 1.2 Summary SECTION 2 DESCRIPTION OF METHODOLOGY 5

2.0 Introduction 2.1 Computer Program Description 2.2 Model Descriptions SECTION 3 DESCRIPTION OF REACTORS USED IN BENCIMARKING 12 3.0 Introduction 3.1 San Onofre Units 2 And 3 SECTION 4 BENCHMARK COMPARISONS 19 4.0 Introduction 4.1 Critical Boron Concentration 4.1.1 Zero Power Critical Boron Concentration 4.1.2 Hot-Full-Power Critical Boron Concentration 4.2 Isothermal Temperature Coefficient 4.3 Power Coefficient 4.4 Control Rod Worth 4.5 Net (N-l) Control Rod Worth 4.6 'Inverse Boron Worth iv

SCE-0901 TABLE OF CONTENTS (Continued)

Page SECTION 5 POWER DISTRIBUTION AND PIN PEAKING COMPARISONS 43 5.0 Introduction 5.1 Local Pin Power 5.2 Pin Peaking Factors (Fq, Fxy, Fr) 5.3 Core Radial And Axial Power Distributions 5.4 Axial Shape Index SECTION 6 FUTURE BURNABLE ABSORBERS AND FUEL ROD CLADDING 53 6.0 Introduction 6.1 Criticals with Gadolinia 6.2 MCNP Comparisons for IFBA and Gadolinia Lattices 6.3 Conclusion For Future Burnable Absorbers 6.4 M5 Cladding SECTION 7 CONCLUSION 62 SECTION 8 REFERENCES 63 V

SCE-0901 LIST OF TABLES Number Title Page 1.1 CASMO-4 / SIMULATE-3 Biases And Uncertainties 4

2.1 CASMO-3 to CASMO-4 Code Feature Comparison 10 3.1 Mechanical Design Parameters - SONGS 2&3 14 4.1 HZP Critical Boron Concentration 28 4.2 SONGS 2 Cycle 12 Critical Boron Concentration - HFP 29 4.3 SONGS 2 Cycle 13 Critical Boron Concentration - BFP 30 4.4 SONGS 2 Cycle 14 Critical Boron Concentration - HFP

,3.1 4.5 SONGS 3 Cycle 12 Critical Boron Concentration - IEP 32!

4.6 SONGS 3 Cycle 13 Critical Boron Concentration - BFP 33 4.7 SONGS 3 Cycle 14 Critical Boron Concentration - BFP 34 4.8 HZP Isothermal Temperature Coefficient - Cycles 10 - 15 35 4.9 HZP Isothermal Temperature Coefficient - Cycles 1 - 5 36 4.10 At-Power Isothermal Temperature Coefficient - Cycles 1 - 5 37 4.11 Power Coefficients 38 4.12 Control Rod Bank Worths - Unit 2 39 4.13 Control Rod Bank Worths - Unit 3 40 4.14 Net (N-1) Control Rod Worth 41 vi

SCE-0901 LIST OF TABLES (Continued)

Number Title Page 4.15 Inverse Boron Worth 42 5.1 Power Distribution Peaking Factor - Fxy 46 5.2 Power Distribution Peaking Factor - Fr 47 5.3 Power Distribution Peaking Factor - Fq 48 vii

SCE-0901 Number 2.1 3.1 3.2 3.3 5.1 5.2 5.3 5.4 LIST OF FIGURES Title Computer Program Sequence Flow Chart SONGS 2 And 3 Reactor Core With Control Rod Pattern SONGS 2 And 3 Reactor Core With Incore Instrument Pattern SONGS 2 And 3 16x16 Fuel Assembly

$2C14 Axially Integrated RPD Comparison 100% RFP -- ARO -- BOC S2C14 Axially Integrated RPD Comparison 100% HFP -- ARO -- EOC S2C14 Average Axial Power Distribution Comparison 100% REP -- ARO -- BOC

$2C14 Average Axial Power Distribution Comparison 100% HFP -- ARO -- EOC Page 11 16 17 18 49 50 51 52 viii

SCE-0901 LIST OF ACRONYMS ARO All Rods Out BOC Beginning Of Cycle CBC Critical Boron Concentration CEA Control Element Assembly (Control Rod)

C-3 CASMO-3 C-4 CASMO-4 C-3/S-3 CASMO-3 / SIMULATE-3 C-4/S-3 CASMO-4 / SIMULATE-3 EARO Essentially All Rods Out EOC End Of Cycle Fr Power Distribution Integrated Peaking Factor Fq Power Distribution Total peaking Factor Fxy Power Distribution Planar Peaking Factor

-FP Hot Full Power HZP Hot Zero Power PCM Percent MNI (Unit of Reactivity = 10i5 Ak/k)

SCE Southern California Edison Company SONGS San Onofre Nuclear Generating Station ix

SCE-0901 SECTION 1 INTRODUCTION, OVERVIEW, AND

1.0 INTRODUCTION

This document describes the Southern California Edison Company (SCE) San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 reactor core physics methodology and model verification using the CASMO-4 and SIMULATE-3 computer programs. Studsvik-Scandpower Inc. is the developer of the CASMO-4 and SIMULATE-3 computer programs.

In June 1992, SCE obtained NRC approval to use CASMO-3 / SIMULATE-3 (Reference 1) for reactor core physics design activities. SCE is upgrading from CASMO-3 (and supporting programs) to CASMO-4 (and supporting programs). The CASMO-4 computer program package is superior in many respects to CASMO-3, which although extremely reliable, has become dated.

This modernization initiative is part of the ongoing commitment of SCE to use best available well established methods for reactor analysis.

1.1 OVERVIEW The data demonstrating the applicability of CASMO-4 for SONGS Units 2 and 3 core physics analyses are documented in Sections 2 through 7.

Section 2, "Description Of Methodology," describes the CASMO-4 / SIMIULATE-3 computer program package.

Section 3, "Description Of Reactors Used in Benchmarking," describes the SONGS reactors

'(Units 2 and 3) and cycles (1 - 5 and 10 - 14) modeled with CASMO-4. San Onofre Unit 1 has been decommissioned.

Section 4, "Benchmark Comparisons," provides the benchmarking of the core physics parameters calculated for SONGS 2&3. CASMO-4 / SIMLULATE-3 results are compared with plant measurements. The sample mean (bias) and standard deviation are quantified. Finally, a 95/95 tolerance limit (Uncertainty) is determined.

Section 5, "Power Distribution Comparisons," provides the benchmarking of the power distribution and pin peaking parameters listed in Table 1.1. It is shown that the power distributions calculated by CASMO-4 / SIM[ULATE-3 and CASMO-3 / SIMULATE-3 are essentially identical. Therefore, the biases and uncertainties for CASMO-3, also apply to CASMO-4.

1

SCE-0901 Section 6, "Future Burnable Absorbers And Fuel Rod Cladding," summarizes critical experiment benchmarking (including gadolinia cores) performed by Studsvik-Scandpower.

MCNP benchmarking results for Integral Fuel Burnable Absorber (IFBA) fuel are also presented. SONGS currently uses erbia burnable absorber but may in the future wish to use gadolinia or JEBA in Lead Fuel Assemblies (LFAs) and then subsequently in full reload batches.

The CASMO-4/SD/IULATE-3 methodology is applicable to these burnable absorbers.

SONGS currently uses Zircaloy-4 and Zirlo fuel rod cladding. AREVA LFAs will have M5 cladding. The CASMO-4 / SIMUJLATE-3 methodology is applicable to Zircaloy-4, Zirlo, and M5 cladding.

Section 7, "Conclusions," presents the conclusions of this report and the range of application for which SCE will use this methodology.

Section 8, "References," presents documents referenced in this report.

2

SCE-0901 1.2

SUMMARY

Table 1.1 summarizes the bias and 95/95 tolerance limits for CASMO-4 / SIMULATE-3 calculated in Sections 4 and 5. The tolerance limits are such that, when applied to the CASMO-4 / SDMULATE-3 results, there is a 95 percent probability, with a 95 percent confidence that the calculated values will conservatively bound the "true" values. Based on the analyses and results contained in this report, the CASMO-4 / SIMULATE-3 methodology may be used for physics analyses of SONGS Units 2 and 3 with the same level of accuracy as CASMO-3.

SCE concludes that CASMO-4 / SIMULATE-3 is acceptable for performing all SONGS Units 2 and 3 steady-state core physics analyses including:

Reload Design Safety Analyses Input Startup Predictions Plant Physics Data Books Reactor Protection and Monitoring System Updates 3

SCE-0901 TABLEJ1.1 CASMO-4 / SIMULATE-3 Biases And Uncertainties Parameter HZP PPM HFP PPM IBW Power Coefficient ITC CEA Bank Worth Local Pin Power Axial Shape Index Assembly Fq Assembly Fxy Assembly Fr Fuel Rod Fq Fuel Rod Fxy Fuel Rod Fr Bias

-1 ppm 30 ppm

-3.9 %

0 pcm/%P

-0.5 pcm/rF 1.3 %

0%

0.003 0.0%

0.0%

0.0 %

0.0%

0.0%

0.0 %

Uncertainty 23 ppm 20 ppm 13.6 %

2 pcxnl%P 2.5 pcmI0F 7.6 %

1.78 %

0.014 4.17 %

4.80 %

3.34 %

4.62 %

5.20 %

3.89 %

Units*

Absolute Absolute Relative \\

Absolute Absolute Relative Relative Absolute Relative Relative Relative Relative Relative Relative

• For those parameters with differences expressed in relative units:

Predicted Value = Calculated Value * [1.0 + (Bias + Uncertainty)/100%]

For those parameters with differences expressed in absolute units:

Predicted Value = Calculated Value + (Bias + Uncertainty) 4

SCE-0901 SECTION 2 DESCRIPTION OF METHODOLOGY

2.0 INTRODUCTION

This section provides a brief description of the CASMO-4 / SIMULATE-3 methodology.

This computer program package has already received NRC approval for use in core physics calculations at other utilities (References 2, 3, 4, 5, and 6).

In general, the methodology is similar to the SCE CASMO-3 Topical Report (Reference 1).

However, the Palo Verde Topical Report (Reference 2) is followed for Inverse Boron Worth and Local Pin Power biases and uncertainties.

2.1 COMPUTER PROGRAM DESCRIPTIONS The CASMO-4 / SIMULATE-3 computer program package consists of the following computer programs developed by Studsvik-Scandpower Inc.:

INTERPIN-3 CASMO-4 CMSLINK SIMULATE-3 The CASMO-4 program package is described below.

CASMO-4 CASMO-4 (Reference 7) is a multi-group, two-dimensional neutron transport theory lattice physics code with depletion capability and the ability to generate cross-sections and discontinuity factors for both boiling water reactor (BWR) and pressurized water reactor (PWR) diffusion theory core analysis. CASMO-4 incorporates the microscopic depletion of burnable absorbers (e.g., gadolinia) directly into the main calculations, and the MICBURN-3 auxiliary code needed by CASMO-3 for gadolinia is no longer required.

One new feature of CASMO-4 is the use of the Method of Characteristics for solving the Boltzmann neutron transport equation, which allows for modeling the true heterogeneous geometry directly in the two-dimensional transport calculation. The Method of Characteristics has been well established as a methodology particularly well suited to these types of problems and is now the solution method of choice for most next generation lattice physics codes:

CASMO-5, HELIOS-2, APOLLO-2, LANCER-02, and WIMS-10.

The CASMO-4 transport method has been well validated in peer reviewed publications (e.g. References 16, 17).

5

SCE-0901 The cross-section generation process with CASMO-4 has also been automated such that all the requisite nuclear data for SIMULATE-3 can be generated in one execution (for one segment).

This automated case matrix feature of CASMO-4 results in reduced potential for human performance error that may arise from the manual construction of the case matrix as was required with CASMO-3. Table 2.1 lists some of the major differences (and similarities) between CASMO-3 and CASMO-4.

As can be seen in Table 2.1, CASMO-4 shares some common heritage with CASMO-3 but also has some key methodological differences that are essentially transparent to the end user, e.g., the 2D transport method implemented. As both codes share the same base ENDF cross section evaluation, built-in biases due to library data should be similar between the two codes.

One of the key similarities between CASMO-3 and CASMO-4 is use of identical input and input preparation which means that training and skills relevant to proficiency at CASMO-3 directly translate into proficiency at CASMO-4. Many of the methodological improvements of CASMO-4 over CASMO-3 are transparent to the end user with the end result that CASMO-4 is intrinsically no more difficult to use than CASMO-3.

CMSLIEK CMSLJNK (Reference 8) is a data processing program that links CASMO-4 to SEVIULATE-3.

The primary purpose of CMSLINK is to read the CASMO-4 ASCII card image file, functionalize key neutronic variables versus important independent variables and produce a binary master cross-section library for use with SIMIULATE-3.

The CMSLINK program processes the following types of data from CASMO-4:

Two-group Macroscopic Cross-sections Assembly Discontinuity Factors (ADF's)

Fission Product Data (fission yields and microscopic cross sections)

Incore Instrument Response Data (detector data)

Pin Power Reconstruction Data Kinetics Data (beta's, lambda's and neutron velocities)

Spontaneous Fission Data/Alpha-n Sources Decay Heat Data Neutronic data is generated by performing CASMO-4 fuel assembly depletions (assuming zero leakage) and branches to off-nominal conditions. These calculations provide all the data needed to functionalize neutronic parameters versus:

Fuel Burnup (EXP)

" Coolant Density or Temperature (DEN)

• Fuel Temperature (TFU)

" Boron Concentration (BOR)

" Control Rod Type (CRD) 6

SCE-0901

• Historical Coolant Density (HDEN)

" Historical Rod Presence (HCRD)

" Historical Boron Concentration (HBOR)

• Historical Fuel Temperature (HTFU)

The data from CASMO-4 is then functionalized by CMSLINK into a series of 1-D, 2-D, and 3-D table sets (for each unique fuel type). This functionalization allows linear interpolation in SIMULATE to accurately evaluate neutronic data for each node in the core for the appropriate reactor conditions.

The functionalizations made in CMSLINK depend upon the amount of detailed branch and depletion data generated in CASMO-4. Normally, several branch cases are made for each independent variable, and simultaneous branches are made for certain variables such as coolant density and control rod. CASMO-4 contains an automated case matrix feature that allows this process to proceed in a standardized fashion.

INTERPIN INTERPIN-3 (Reference 9) is used to calculate fuel pin temperatures. SCE uses this computer program to calculate the fuel temperature of the average rod as a function of bumup. Output from this computer program provides a single burnup independent fuel rod temperature for use in CASMO-4, and a power and b umup dependent fuel rod temperature for use in S]VIULATE-3.

SIMULATE-3 SJMIULATE-3 (Reference 10) is a two-group, three-dimensional (3-D), coarse mesh nodal diffusion theory reactor simulator computer program that employs both thermal-hydraulic and Doppler feedback. In the axial direction, 12-25 nodes are typically used to represent the active portion of each fuel assembly, and one node is used to represent the upper and lower reflectors.

SIMULATE-3 explicitly models the baffle/reflector region, eliminating the need to normalize to higher-order fine mesh calculations such as PDQ (a diffusion theory computer program).

Homogenized cross-sections and discontinuity factors are applied to the coarse mesh nodal model to solve the two-grouip diffusion equation using the QPANDA neutronics model.

QPANDA employs fourth order polynomial representations of the intra-nodal flux distributions in both the fast and thermal groups.

The nodal thermal-hydraulic properties are calculated based on the inlet temperature, RCS pressure, coolant mass flow rate, and the heat addition along the channel.

The pin-by-pin power distributions, on a 2-D or 3-D basis, are constructed from the inter-and intra-assembly information from the coarse mesh solution and the pin-wise assembly power distribution from CASMO-4.

7

SCE-0901 SIEMULATE-3 performs a macroscopic depletion and individual uranium, plutonium, and fission product isotope concentrations are not computed. However, microscopic depletion of iodine, xenon, promethium, and samarium is included to allow modeling of typical reactor transients.

2.2 MODEL DESCRIPTIONS CASMO-4 Fuel Assembly And Reflector Models Each unique SONGS fuel assembly type (defined by geometry, enrichment, and burnable poison) is separately modeled in CASMO-4 using octant symmetry. Enrichment zoning among fuel rods, burnable absorbers (separate rods and mixed with the U0 2 fuel) and guide tubes are explicitly modeled. Any water gap between fuel assemblies in the core is included in the CASMO-4 model. The spacer grids are also modeled. Design basis documents such as the SONGS Reload Ground Rules and as-built drawings provide the necessary data to develop the CASMO-4 assembly models.

Several depletions are typically needed to generate each fuel assembly type's cross-section'data.'.

First, the fuel assembly is depleted at hot full power, reactor average conditions (base depletion).

Moderator temperature, fuel temperature, and. soluble boron concentration are set to constant.

average values for the complete depletion. The average fuel temperature at hot full power conditions is calculated with INTERPIN-3. Next the fuel assembly is depleted at a lower moderator temperature, typically TMO-20 K, (moderator temperature history depletion).

However, the fuel temperature and the soluble boron concentration are kept at the constant hot full power, reactor average values. The fuel assembly is again depleted at constant hot full power, reactor, average conditions, but with a constant soluble boron concentration higher than is usually seen in normal operation, typically 2*BOR where BOR is the boron at base conditions, (boron history depletion). A fuel temperature history depletion is then performed at TFU=TMO-20 K, and finally an optional rodded depletion can be performed. Each fuel assembly is depleted up to 60 GWD/T assembly average burnup using the CASMO-4 default depletion steps.

Branch cases are performed to calculate instantaneous effects. These instantaneous effects are individually calculated and added together later to recreate the proper fuel assembly cross-sections. The branch cases are executed from the hot full power reactor average conditions depletion case at typically 0, 5, 10, 20, 30, 40, 50, and 60 GWD/T. Branch cases are run for off-nominal moderator temperatures, fuel temperatures, soluble boron concentrations, and control rod insertions to encompass the full range of conditions expected during reactor startup and operation.,

A single CASMO-4 case performs all of the above depletions and branch cases. Furthermore, if gadolinia is the burnable absorber, no auxiliary program is needed as the depletion of gadolinia is modeled directly in CASMO-4.

CASMO-4 also generates explicit top, bottom, and radial reflector cross-sections (which are important for PWR modeling). The radial reflector consists of the stainless steel core baffle 8

SCE-0901 followed by about 15 centimeters (cm) of water. The top reflector extends from the top of the active fuel to the lower surface of the fuel assembly upper end fitting. The bottom reflector extends from the bottom of the active fuel to the lower surface of the core support plate.

Reflector cross-sections are typically modeled as a function of soluble boron concentration and moderator temperature. CASMO-4 performs reflector calculations in 25 energy groups to accurately capture the effects of high energy leakage.

CMSLINK Model The CMSLINK computer program generates two-group fuel assembly and reflector cross-section tables for SNI-ULATE-3. For each fuel assembly type and reflector type, data from the CASMO-4 card-image file are processed into a binary cross-section library for input to SIMULATE-3.

SIMULATE-3 Model The SIMULATE-3 model typically divides the active fuel region into 20 axial nodes and four radial nodes per fuel assembly. A pseudo-assembly, consisting of reflector material, surrounds the core and is divided into one radial and 20 axial nodes. Axially, the fuel is divided into a single.bottom reflector node, 20 active fuel nodes, 'and a single top reflector node.

Additional model input data are:

Full Core Fuel Assembly Serial Number Map Quarter Core Fuel Assembly Type Map Fuel Assembly Axial Zone Definition, Including Reflectors Cross-Section Library Assignment To Fuel Assembly And Reflector Types Control Rod Locations Grouping of Control Rods Into Banks Axial Zone Definitions For Control Rods, especially Part Length Control Rods In-Core Instrument Locations Fuel Temperature versus Power and Burnup Correlation (INTERPIN-3 Program)

Core Mw-Thermal Output At 100% Power Core Pressure And Coolant Mass Flow Rate At 100% Power Core Inlet Temperature Versus Power Level Input Restart File (Cycle N-i) And Cross-Section Library File Output Restart File After the cycle base model is set up, the physics analyst can specify the percent power level, control rod bank positions (percent withdrawn), output and edit options, and the type of calculation: depletion, xenon transient, coefficient calculation (e.g. ITC, IBW, FTC, etc.).

9

SCE-0901 TABLE 2.1 CASMO-3 to CASMO-4 Code Feature ComDarison CASMO-3 CASMO-4 Differences between CASMO-3 and CASMO-4 Library Data Evaluation ENDF/B-IV ENDF/B-IV

1. of energy groups in neutron data library 40 or 70 70

1. of nuclides in neutron data library 93 103 2D Transmission Method of transport method probabilities Characteristics Geometry homogeneous heterogeneous Gd depletion External code Internal depletion Case Matrix Manual Input Automated Case Matrix Default # 2D energy groups (non-Ref) 7 8

Default # 2D energy groups (Ref) 7 25 Similarities between CASMO-3 and CASMO-4 Pin Cell Calculation Collision Probabilities Resonance Calculation Equivalence Theory Burnup Calculation Predictor/Corrector Simple Engineering Data Input

-Identical for CASMO-3 and CASMO-4 Thermal Expansion Yes Generates 2 Group Data for SIMULATE-3 Yes 10

SCE-0901 FIGURE 2.1 Computer Program Seauence Flow Chart INTERPIN V

v CASMO-4 V

CMSLINK v

INTERPIN --------

> SIMULATE-3 v

Assembly Power Distribution Pin-By-Pin Power Distribution Critical Boron Concentration Cycle Length Reactivity Coefficients Etc.

11

SCE-0901 SECTION 3 DESCRIPTION OF REACTORS USED IN BENCHMARKING

3.0 INTRODUCTION

This report compares the CASMO-4 / SIMULATE-3 predictions of key physics parameters with measured data and CASMO-3 / SIMULATE-3 results. The reactor plants are San Onofre Nuclear Generating Station Units 2 and 3. The measured data were obtained during plant startup and normal operation.

The following Section provides a brief description of San Onofre Units 2 And 3. Detailed information can be found in the SONGS Units 2 and 3 UFSAR (Reference 11).

3.1 SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 (SONGS 2&3)

SONGS 2&3 are commercial nuclear power plants. SONGS 2 began commercial operation in 1983. SONGS 3 began commercial operation in 1984. SONGS 2 is in its 1 5 th cycle of operation.

SONGS 3 is in its 15'h cycle of operation. SONGS 2&3 are Combustion Engineering (absorbed into Westinghouse) two-loop pressurized water reactors (PW-Rs). Each unit produces 3438-megawatts-thermal at 100% rated power. (The power has been: uprated from 3390 Mw-t.)

Each reactor core contains 217 fuel assemblies arranged as shown in Figure 3.1. Both in-out and, low-leakage fuel management patterns have been used. Each fuel assembly consists of a 16x16 array of 236 fuel rods and 5 control rod guide tubes (Figure 3.3). Core, fuel assembly, control rod, and burnable absorber data are summarized in Table 3.1,.

The fuel rods are low enriched (< 4.8 w/o) U0 2 pellets clad in Zircaloy-4 or Zirlo. The control rod guide tubes are also Zircaloy-4. Ten Zircaloy-4 grids and one Inconel-718 grid are located along the length of the assembly.

The incore instrumentation system for power distribution measurement consists of 56 strings of fixed Rhodium detectors (Figure 3.2). Each detector string consists of five individual, 40 cm long, Rhodium detectors placed at about 15, 30, 50, 70, and 90 percent of active core height.

The detector signals are processed of-line with the Combustion Engineering (now Westinghouse)

CECOR (Reference 13) computer program to determine the power distribution in the core.

There are 83 full-length and eight part-length control rods, called control element assemblies (CEA's). Seventy-nine full length CEA's have 5 individual absorber rods. Four full-length CEA"s located on the core periphery have 4 individual absorber rods. The full length CEA's have an Inconel nose cap, about 6" of AgInCd, and about 136" of B4C pellets. The eight part length CEA's have five absorber rods consisting of approximately 75" of Inconel, 58" of water 12

SCE-0901 filled Inconel tube, and 16" of B 4C pellets. The cladding material is Inconel-625. The CEA's are moved in nine symmetrical groups:

Regulating Groups 1 Through 6 Part Length Group Shutdown Groups A and B Burnable absorber rods, consisting of B4C-A120 3 pellets in Zircaloy-4 cladding, were used in Cycles 1 through 8 of both Units. Lead Fuel Assemblies (LFA's) with erbia (Er2O3) mixed with the U0 2 in the fuel rod were placed in the SONGS 2 reactor core during cycle 6. Cycle 9 of each Unit was a transition cycle consisting of a full, fresh reload batch with erbia as the burnable absorber, and a burned batch with depleted B4C-A120 3 absorber rods. Cycles 10 through the present have used erbia as the burnable absorber. Future cycles may employ:

(1) Gadolinia (Gd20 3) mixed with U0 2 in the fuel rod with Zircaloy-4, Zirlo, or M5 cladding, and/or, (2) Integral Fuel Burnable Absorber (IFBA) - A ZrB2 coating on the surface of the U02 fuel pellet in Zircaloy-4, Zirlo, or M5 cladding.

The SONGS 2&3 reactors have several unique features. The outermost row of four fuel assemblies does not line up with the next interior row of fuel assemblies. The four-finger CEA inserted in the middle pair of these "off-set" assemblies has two fingers in one assembly and two fingers in the adjacent assembly. The B4C-A120 3 burnable absorber rods and erbia burnable absorber do not extend the full length of the active fuel region resulting in axially zoned fuel assemblies. [Future burnable absorbers such as Gadolinia or IFBA would also be axially zoned.] The cycle length of both Units has increased from 366 Effective Full Power Days (EFPD) for Cycle 1 to about 600 EFPD for Cycle 15. Finally, the five control rod guide tubes are large (displacing 4 fuel rods) compared to Westinghouse fuel (displacing 1 fuel rod) and Babcock & Wilcox fuel (displacing 1 fuel rod) fuel assembly designs.

13

SCE-0901 TABLE 3.1 Mechanical Design Parameters SONGS 2&3 Core Description Power Level Number of Assemblies Number Of Control Rods Fuel Assembly Pitch Core Area Core Equivalent Diameter Fuel Assembly Description 3390 / 3438 Megawatts-Thermal 217 91 8.180 inches 101.1 Square Feet 136 inches Fuel Rod Array Fuel Rod Pitch Outside Dimension Number Of Guide Tubes Guide Tube I.D.

Guide Tube O.D.

Guide Tube Material Fuel Rod Description Material Maximum Enrichment Stack Height Density Pellet Diameter Clad Material Clad I.D.

Clad O.D.

Clad Thickness Active Fuel Length 16 x 16 0.506 inches 7.972 inches 5

0.90 inches 0.98 inches Zircaloy-4 U0 2 4.8 w/o U235 10.06 -- 10.34 gm/cm 3 0.3250 / 0.3255 inches Zircaloy-4 and Zirlo 0.332 inches 0.382 inches 0.025 inches 150 inches Full-Lenath Control Rod Descriution Number 5-Finger 4-Finger Clad Material Clad Thickness Clad O.D.

83 79 4

Inconel-625 0.035 inches 0.816 inches 14

SCE-0901 TABLE 3.1 (Continued)

Full-Length Control Rod (Continued)

Poison Material Poison Material Length 5-Finger 4-Finger B4C Pellet Diameter

% T. D. Of 2.52 g/cm 3 Weight % Boron, Minimum Part Length Control Rod Description Number Clad Material Clad Thickness Clad O.D.

Poison Material Poison Material Length B 4C Pellet Diameter

% T. D. Of 2.52 g/cm 3 Weight % Boron, Minimum B4C / Ag-In-Cd / Inconel 136.0" 12.5" 126.5" 12.5" 0.737 inches 73 %

77.5 %

0.6" 10.1" 8 (5-fingers)

Inconel-625 0,035 inches 0.816 inches Inconel / Water / B 4 C 76.4" 55.0" 16.0" 0.737 inches 73 %

77.5 %

Burnable Poison Descriution Absorber Material U0 2 - Er20 3 T.D.

Number Per Fuel Assembly Er2O3 10.3 gMlcm 3 0, 40, 48, 60, 72, and 80 Absorber Material Pellet Diameter Pellet Length Active Length Clad Material Clad I.D.

Clad O.D.

A120 3 - B4C 0.310 inches 0.50 inches min 136 inches Zircaloy-4 0.332 inches 0.382 inches 15

SCE-0901 FIGURE 3.1 SONGS 2 And 3 Reactor Core With Control Rod Pattern AA Banks 1 - 6 are Regulating Banks.

Banks A and B are Shutdown Banks.

(AA Indicates a 4-Finger Control Rod spanning Two Assemblies on the core periphery.)

Bank P is the Part-Length Control Rods.

A*: If the quadrant shown is #1, this control rod is not present in quadrants #2 and #4.

16

SCE-0901 FIGURE 3.2 SONGS 2 And 3 Reactor Core With Incore Instrument Pattern

-X = Incore Instrument Location 17

SCE-0901 FIGURE 3.3 SONGS 2 And 3 16x16 Fuel Assembly x

x xxxxxxx xxxxxx XXX x x x x x x x x xxx xxx xxx X X x

x x

xxxxx xxx xxx x x x x x x x x x x x x x x x x x x x x x x x x x x x x

xxxxx x

xxxxxx x

xxxxxx xxxxxx xxxx x x x x x x x x xxxx xxxx xxxx X X x

x x

xxx xxxxxxxxxx x

x xxx xxxxxxxx xx x

x xxx xxxxxxxx xx XXX XXX XXX XXX xxx XXX XXX xxx xxx XXX XXX xxx XXX XXX XXX x

x x

xxxx x

xxx xx x

xxxxx x

xx xx xxxxxxx xxx xxx xxx xxx XXX xxx xxx xxx xxx x x

xxx XXX xxx XXX xxx xxx xxxxx x x x

x x x

x x x

x x x

xxxxx xxxxx x xxxx XX xX x

xxx x x x x x x x x x x x x x x x x x x x x x x X; xx xx 7.972"

<K------------------------

0.208" 7.972" 8.180" X = Fuel Rod There are 5 control rod guide tubes. Each guide tube displaces 4 fuel rods.

Pellet Diameter Clad I.D.

Clad O.D.

Fuel Rod Pitch Active Fuel Length Guide Tube I.D.

Guide Tube O.D.

0.3255 inches 0.322 inches 0.382 inches 0.506 inches 150 inches 0.90 inches 0.98 inches I

18

SCE-0901 SECTION 4 BENCHMARK COMPARISONS

4.0 INTRODUCTION

This section compares CASMO-4 / SIMULATE-3 calculated parameters with measured plant data. The measured data are from zero power startup testing, at-power Isothermal Temperature Coefficient measurements, and normal operations at SONGS 2&3. For each parameter compared, the sample mean and standard deviation of the observed differences are calculated.

Based on the mean, standard deviation, and sample size, a bias and conservative 95/95 tolerance limit (bias + uncertainty) are calculated.

Typically for this benchmarking, SONGS 2 and 3 cycles that have erbia as the burnable absorber and are most representative of the future and current fuel management strategy (checkerboard) and cycle length (- 600 EFPD) will be used. These cycles are SONGS 2 Cycles 10 - 15 and SONGS 3 Cycles 10 - 14. However, the power coefficient has not been measured in these recent cycles. The power coefficient was measured in Cycles 1 - 3. Therefore, the Cycles 1 - 3 measurements will be analyzed with CASMO-4. The ITC measurement from Cycles 10- 15.

generally occurred around 2000 ppm. To obtain a greaterrange of soluble boron for the ITC measurements, Cycles 1-5 ITC measurements. will be analyzed with CASMQ-4.

Section 4.1 presents the Critical Boron Concentration (0CBC) omparisons for zero power and hot full power conditions. Differences between calculated and measured data are represented in absolute terms (Measured - Calculated).

Section 4.2 presents the Isothermal Temperature Coefficient (ITC) comparisons for zero power and hot full power conditions. Differences between calculated and measured data are represented in absolute terms (Measured - Calculated).

Section 4.3 presents the Power Coefficient (PC) comparisons. Differences between calculated and measured data are represented in absolute terms (Measured - Calculated).

Section 4.4 presents the Control Rod Worth comparisons. Differences between calculated and measured data are represented in relative terms - 100% * (Measured - Calculated) / Calculated.

19

SCE-0901 Section 4.5 verifies the ability of CASMO-4 / SIMULATE-3 to predict the net (N-I) control rod worth. This section is not a comparison against measured data. CASMO-4 / SIIMULATE-3 results are compared against CASMO-3 / SIMULATE-3 results.

Section 4.6 presents the Inverse Boron Worth (1BW) comparisons. Differences are represented in relative terms.

20

SCE-0901 4.1 CRITICAL BORON CONCENTRATION CASMO-4 1 SIMULATE-3 Critical Boron Concentrations (CBC) were compared to zero-power startup measurements' and full power operating data. The most reliable measurements are the zero-power startup tests. These measurements are made under well controlled conditions without significant thermal and xenon feedbacks, and no boron depletion effects. The zero-power comparison statistics quantify SIMULATE-3's accuracy in predicting CBC for Beginning-Of-Cycle (BOC), zero-power conditions without xenon in the core.

The full-power operation boron concentration data are from titration of reactor coolant samples.

All measurements are adjusted for control rod insertions and deviations from full-power equilibrium conditions. These full power comparisons serve as conservative estimates of the SIHMULATE-3 uncertainties for at-power equilibrium conditions with thermal feedback.

Sections 4.1.1 and 4.1.2 present the comparisons for zero-power and full-power CBC.

4.1.1 ZERO-POWER CRITICAL BORON CONCENTRATION Table 4.1 lists the measured and SIMULATE-3 predicted values for BOC, zero-power, xenon free Critical Boron Concentrations (CBC). Eleven measurements from the most recent eleven cycles of startup tests have been included. All measurements are at about 545 degrees F and' essentially all control rods out.

A three-step statistical analysis was performed orivthe measured and SIMULATE-3 calculated CBC differences. First, the sample mean (x-bar), and standard deviation (S) were calculated for the CBC differences. The differences are due to SINMULATE-3 calculational uncertainties, variations in B-10 isotopic concentrations, and measurement (titration) uncertainties. For example, boron concentration measurement errors can be as high as 5 ppm. For conservatism, all differences are assumed due only to SIIMULATE-3 calculational uncertainties.

Second, the sample distribution of differences is tested for normality using ANSI N15.15-1974 (Reference 14). The normality test is needed because the 95/95 tolerance limit assumes that the sample differences are normally distributed.

Finally the bias and 95/95 tolerance limit (Uncertainty) are calculated. The bias is equal to the sample mean (x-bar). The 95/95 tolerance limit is calculated by K95/95

• S, where S is the standard deviation and K95/95 is determined from Reference 15.

Calculation of the CASMO-4 / SIMULATE-3 zero power CBC bias and 95/95 tolerance limit is shown in Table 4.1:

-1 ppm + 23 ppm 21

SCE-0901 4.1.2 HOT-FULL-POWER CRITICAL BORON CONCENTRATION For this benchmarking effort, the most recent six cycles (SONGS 2 and 3 Cycles 12 - 14) were determined to be sufficient. There are a total of 111 comparisons. Therefore, Cycles 10 and 11I were not analyzed.

Tables 4.2 through 4.7 compare Cycles 12 - 14 CASMO-4 / SIMULATE-3 calculated Hot-Full-Power (I-FP) CBCs with measured data taken during cycle operation.

The SIMULATE-3 HFP CBC and 95/95 tolerance limits were determined using the statistical methods outlined in Section 4.1.1.

Calculation of the CASMO-4 / SIMULATE-3 full power CBC bias and 95/95 tolerance limit is shown in Table 4.7:

30 ppm + 20 ppm 22

SCE-0901 4.2 ISOTHERMAL TEMPERATURE COEFFICIENT The Isothermal Temperature Coefficient (ITC) is the change in the reactivity due to a 1 degree F change in the core average moderator and fuel temperature. Tables 4.8, 4.9, and 4.10 list the comparisons of zero-power and at-power calculated ITC's with measurements.

A statistical analysis has been performed on the ITC differences using the process outlined in Section 4.1.1.

4.2.1 HZP ITC Calculation of the CASMO-4 / SIMNULATE-3 zero-power ITC bias and uncertainty for the HZP ITC measured in SONGS 2 Cycles 10 - 15 and SONGS 3 Cycles 10-14 is shown in Table 4.8:

-0.15 pcm/°F + 0.36 pcm/PF 4.2.2 HZP AND AT-POWER ITC CYCLES 1 - 5 The ITC measurement from Cycles 10 - 15 generally occurred around 2000 ppm. To obtain a greater range of soluble boron for the ITO measurements, Cycles 1-5 ITC measurements will be analyzed with CASMO-4.

Calculation of the CASMO-4 / SIMULATE-3 bias and uncertainty for the HZP and at-power ITC's measured in Cycles 1 - 5 is shown in Tables 4.9 and 4.10:

-0.46 pcmI°F + 2.50 pcm/F 4.2.3 FINAL HZP AND HFP ITC In order to expand the range of applicability, results from Cycles 1 - 5 and Cycles 10 - 15 were combined. The combined Cycles 1 - 5 and Cycles 10 -15 results are not a normal distribution.

However, the Cycles 1 - 5 results provide a conservative bias and uncertainty which covers all HZP and P-P conditions at a wide range of soluble boron concentrations.

Therefore, the final bias and uncertainty are: -0.5 + 2.5 pcm / 'F 23

SCE-0901 4.3 POWER COEFFICIENT The power coefficient is defined as the change in reactivity due to a change in the core power level. Since the power coefficient has not been measured for recent cycles, CASMO-4 /

SIMULATE-3 power coefficient predictions were compared to measurements from Cycles 1, 2, and 3, and summarized in Table 4.11.

Due to the limited size of the database, a meaningful 95/95 tolerance limit could not be derived.

However, all of the differences are within 2 pcm/%P, and the sample mean and standard deviation are -0.55 pcm/%P and 0.78 pcm/%P, respectively. Since the differences include both the calculational and the measurement uncertainties, a conservative 95/95 tolerance limit of 2 pcmr%P can be assumed based on sound engineering judgment.

Therefore, the CASMO-4 / SIMULATE-3 Power Coefficient bias and 95/95 tolerance limit is:

0 pcrrl%P + 2 pcmr%P 24

SCE-0901 4.4 CONTROL ROD WORTH Tables 4.12 and 4.13 list the measured and the calculated control rod bank worths.

The bank worths were measured during startup testing at BOC at zero-power conditions.

Fifty-two (52) measurements from 10 cycles are shown. The STAR (Startup Test Activity Reduction) program (Reference 24) was implemented beginning with Unit 2 Cycle 15.

Therefore, the Control Rod Bank worths were not measured in Cycle 15.

A statistical analysis has been performed on the control rod bank worth differences.

The analysis determined the bias (mean), standard deviation, and normality of the difference distribution. The mean is 1.3% and the standard deviation is 4.67%.

The measurement uncertainty has two components: the measurement uncertainty and the calculational uncertainty. These two components are related to the observed uncertainty by:

SOBS 2

= SM 2 +

Sc 2

The measurement uncertainty has been quantified in Reference 1. Since SONGS 2&3 Cycle 1 were duplicate plants (identical fuel management, enrichments, burnable absorber worth,ýetc) one would expect the measured control rod worths at the beginning of the first cycle to be exactly the same. Therefore, the observed differences in Cycle 1 measurements is attributable to the measurement uncertainty. In Reference 1, Section 4.4, the standard deviation (SD) of the differences in SONGS 2&3 Cycle 1 measured control rod worths'was determined to be 4.0 %.

Therefore, the net measurement uncertainty is:

Measurement Uncty = SM2 = 1/2

• SD2 = 8.0%

Once the measurement uncertainty is quantified, ýthe control rod worth calculational uncertainty is:

Sc

[SOBS SM2 ]1/2

= [(4.67)2 - (8.00) ]1/2

= 3.72 %

Finally, the 95/95 tolerance limit is: 95/95 Tolerance Limits = K95/95

• SC Substituting the appropriate values into the above formula as shown in Table 4.13, the CASMO-4 / SIMJULATE-3 control rod worth bias and 95/95 tolerance limit is:

1.3%

+ 7.6%

The bias and tolerance limit will be applied to the SIMELATE-3 calculation of control rod worths at all power and moderator temperature conditions.

25

SCE-0901 4.5 NET (N-1) CONTROL ROD WORTH The net (N-i) control rod worth is defined as the reactivity worth of the insertion of all of the control rods except the most reactive control rod, which remains stuck out. Due to the high peaking in the assembly in which the stuck control rod is located, this configuration represents one of the most severe challenges to any reactor physics analysis method.

CASMO-3 / SIMULATE-3 capabilities in predicting the net control rod worth and the worst stuck rod worth was verified by predicting the measurement performed during the initial startup of Arkansas Nuclear One - Unit 2 (ANO-2) (Reference 1, Section 4.5).

The CASMO-3 / SIMULATE-3 prediction compared well with the measured data. It was concluded that the bias and 95/95 tolerance limit for control rod bank worth (Section 4.4) was applicable to the net (N-i) worth also.

In Table 4.14, CASMO-3 /SIMULATE-3 and CASMO-4 /SIMULATE-3 results for total worth, stuck rod worth, and (N-i) worth for SONGS Unit 2 Cycle 14 are compared. Differences are less than 1.4% and well within the 95/95 tolerance limit for control rod bank worth (Section 4.4).

Therefore, it is concluded that the bias and 95/95 tolerance limit for control rod bank worth (Section 4.4) is applicable to the net (N-l) worth also.

26

SCE-0901 4.6 INVERSE BORON WORTH The Inverse Boron Worth (IBW) is defined as one over (the inverse) the reactivity worth of the soluble boron dissolved in the reactor coolant - ppm / % Ap.

The measured IBW is calculated using:

E13W = -(CBCj - CBC 2) / (%AP from Control Rod Insertion / Withdrawl) where CBCa, CBC 2, and %APcontrol Rods are measured.

The measurement uncertainty includes boron titration errors and control rod worth measurement errors.

Calculated and measured Inverse Boron Worths are compared in Table 4.15.

A mean (bias) and standard deviation have been calculated based on all the data.

The distribution is normal. Finally, a 95/95 tolerance factor was determined.

The CASMO-4 / SIMULATE-3 IBW bias,and uncertainty are:

-3.9 % + 13.6 %

NOTES: (1) The IBW measurement in Unit 3 Cycle 12 was identified as bad and thus not included in the benchmarking.

(2) With implementation of the STAR program (Reference 24), the IBW was not measured in Cycle 15.

27

SCE-0901 TABLE 4.1 HZP Critical Boron Concentration Beginning Of Cycle, Essentially All Rods Out Unit Cycle Measurement (ppm) 2 2

2 2

2 2

3 3

3 3

3 10 11 12 13 14 15 10 11 12 13 14 1992 1936 2043 2065 2063 2148 2018 2052 2115 2100 2093 C-4/S-3 Prediction (ppm) 2005 1935 2034 2061 2071 2139 2021 2044 2127 2105 2091 Bias =

Difference (ppm)*

-13 1

9 4

-8 9

-3 8

-12

-5.

2

-1

• Measurement - Prediction CASMO-4 Uncertainty --

Normality Test Result = Normal Sample Size = 11 Degrees Of Freedom = 10 K95195 = 2.815 Standard Deviation (S) = 8 ppm 95195 Uncertainty (K 95/95

• S) = 23 ppm 28

SCE-0901 TABLE 4.2 SONGS 2 Cycle 12 Critical Boron Concentration -- -IFP MEASURED SIMULATE-3 M-S3 Burnup Power EFPD (GWD/T)

PPM PPM Delta-PPM S2C12 100 20.35 0.738 1505 1477 28 100 47.33 1.716 1463 1432 31 100 75.13 2.724 1418 1384 34 100 103.90 3.767 1360 1325 35 100 127.49 4.622 1293 1271 22 100 157.34 5.704 1220 1198 22 100 183.27 6.644 1160 1131 29 100 234.39 8.498 996 991 5

100 265.25 9.616 914 903 11 100 292.11 10.590 835 825 10 100 319.97 11.600 759 742 17 100 373.40 13.537 614 580 34 100 401.26 14.547 523 496 27 100 429.09 15.556 440 410 30 100 456.88 16.564 359 325 34 100 512.83 18.592 193 154 39 100 540.78 19.605 106 70 36 100 568.73 20.619 22

-14 36 29

SCE-0901 TABLE 4.3 SONGS 2 Cycle 13 Critical Boron Concentration -- HIFP MEASURED SIMULATE-3 M - S3 Burnup Power EFPD (GWD/T)

PPM PPM Delta-PPM S2C13 loo 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 26.0 0.942 53.8 1.949 84.7 3.068 114.2 4.136 145.3 5.263 179.3 6.494 206.2 7.469 230.0 8.331 264.7 9.587 290.7 10.529 320.0 11.590 348.0 12.605 373.5 13.528 401.2 14.531 428.9 15.535 456.7 16.542 484.4 17.545 512.0 18.545 1499 1444 1386 1330 1253 1169 1087 1032 928 857 756 685 608 530 443 358 274 188 1466 1419 1360 1297 1223 1136 1064 998 899 824 738 654 577 493 408 323 239 156 33 25 26 33 30 33 23 34 29 33 18 31 31 37 35 35 35 32 30

SCE-0901 TABLE 4.4 SONGS 2 Cycle 14 Critical Boron Concentration -- HFP MEASURED SIMULATE-3 M - S3 Burnup Power EFPD (GWDiT)

PPM PPM Delta-PPM S2C14 100 44.07 1.595 1439 1430 9

100 64.73 2.343 1406 1396 10 100 92.33 3.342 1357 1341 16 100 99.23 3.592 1344 1326 18 100 127.01 4.598 1281 1263 18 100 154.73 5.601 1211 1195 16 100 182.46 6.605 1142 1123 19 100 216.52 7.838 1053 1031 22 100 244.97 8.868 975 951 24 100 279.34 10.112 879 853 26 100 307.12 11.118 786 771 15 100 334.86 12.122 709 689 20 100 362.65 13.128 629 605 24 100 397.26 14.381 524 501 23 100 415.75 15.050 470 445 25 100 441.03 15.965 394 368 26 100 475.40 17.209 290 263 27 100 503.10 18.212 212 179 33 100 523.91 18.966 145 117 28 100 556.80 20.156 47 19 28 31

SCE-0901 TABLE 4.5 SONGS 3 Cycle 12 Critical Boron Concentration -- HFP MEASURED SIMULATE-3 M-S3 Burnup Power EFPD (GWD/T)

PPM PPM Delta-PPM S3C12 100 43.23 1.567 1499 1487 12 100 77.12 2.796 1447 1427 20 100 104.00 3.771 1395 1371 24 100 134.83 4.889 1326 1298 28 100 166.90 6.051 1241 1217 24 100 194.87 7.065 1158 1143 15 100 222.83 8.079 1089 1065 24 100 252.90 9.169 1011 979 32 100 281.55 10.208 923 895 28 100 314.29 11.395 815 797 18 100 334.27 12.120 756 737 19 100 369.49 13.397 659 628 31 100 406.41 14.735

.556 514 42 100 434.39 15.750 474 427 47 100 464.79 16.852 374 333 41 100 493.78 17.903 299 243 56 100 522.71 18.952 213 154 59 100 549.55 19.925 133 73 60 32

SCE-0901 TABLE 4.6 SONGS 3 Cycle 13 Critical Boron Concentration -- HEFP MEASURED SIMULATE-3 M - S3 Burnup Power EFPD (GWDfT)

PPM PPM Delta-PPM S3C13 100 16.2 0.587 1554 1524 30 100 43.3 1.570 1501 1475 26 100 71.3 2.585 1461 1425 36 100 99.1 3.593 1398 1368 30 100 127.7 4.629 1322 1301 21 100 148.7 5.391 1287 1249 38 100 176.6 6.402 1213 1177 36 100 204.8 7.424 1132 1100 32 100 232.5 8.429 1060 1023 37 100 260.5 9.444 980 942 38 100 288.5 10.459 901 860 41 100 316.3 11.467 820 777 43 100 343.8 12.463 740 694 46 100 371.8 13.478 651 608 43 100 399.8 14.494 568 523 45 100 427.7 15.505 482 437 45 100 455.3 16.506 397 352 45 33

SCE-0901 TABLE 4.7 SONGS 2 Cycle 14 Critical Boron Concentration -- HFP Burnup Power EFPD (GWD/T MEASURED PPM

.SIMULATE-3 PPM M - S3 Delta-PPM S3C14 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 100 7.86 0.285 39.95 67.94 95.83 123.76 156.57 184.73 212.54 247.43 275.39 306.49 334.47 369.40 396.01 423.89 437.90 456.57 484.22 519.15 546.00 1.448 2.463 3.474 4.486 5.676 6.696 7.704 8.969 9.983 11.110 12.124 13.390 14.355 15.366 15.873 16.550 17.552 18.819 19.792 1549 1466 1421 1370 1311 1228 1159 1089 993 917 821 737 634 556 474 429 384 290 188 107 1509 1444 1397 1341 1279 1199 1127 1052 956

.876 786 703 599 519 434 392 336 252 147 67 40 22 24 29 32 29 32 37 37 41 35 34 35 37 40 37 48 38 41 40 Bias from 111 measurements = 30 ppm CASMO-4 Uncertainty -- Normality Test Result = Normal Sample Size = 111 Degrees Of Freedom = 110 K95/95 = 1.911 Standard Deviation (S) = 10.3 ppm 95/95 Uncertainty (K95/95 *S) = 20 ppm 34

SCE-0901 TABLE 4.8 HZP Isothermal Temperature Coefficient - Cycles 10 - 15 Beginning of Cycle, Essentially All Rods Out Unit Cycle 2

2 2

2 2

2 3

3 3

3 3

10 11 12 13 14 15 10 11 12 13 14 Boron (ppm) 1992 1931 2035 2060 2063 2144 2020 2047 2118 2100 2090 Measurement (pcm/'F)

-1.84

-2.16

-1.69

-1.44

-1.41

-1.16

-1.98

-1.79

-1.09

-1.42

-1.72 C-4/S-3 Prediction (pcm/F)

-1.81

-2.06

-1.63

-1.53

-1.10

-0.84

-1.88

-1.66

-0.91

-1.19

-1.44 Bias =

M - S3 (pcm/IF)

-0.03

-0.10

-0.06 0.09

-0.31

-0.32

-0.10

-0.13

-0.18

-0.23

-0.28

-0.15 CASMO-4 Uncertainty --

t Normality Test Result = Normal Sample Size = 11 Degrees Of Freedom = 10 K95/95 = 2.815 Standard Deviation (S) = 0.13 pcm/F 95/95 Uncertainty (K95/95

• S) = 0.36 pcmfF 35

SCE-0901 TABLE 4.9 HZP Isothermal Temperature Coefficient - Cycles 1 - 5 Beginning of Cycle MEAS CASMO4 M-C4 Unit Cycle CBC Temp CEA ITC ITC ITC (PPM)

(Deg F)

Position (pcm/IF)

(pcm/°F)

(pcm/F) 2 1

869 320 ARO

-1.43

-1.90 0.47 797 320 Bk 6-4 In

-3.46

-4.26 0.80 833 545 ARO

-3.80

-4.21 0.41 2

2 1198.,

545 ARO 0.75 2.65

-1.90 883 545 Bk 6-1 In

-9.14

-7.48

-1.66 2

3 1580 545 ARO 0.50 1.97

-1.47 1382 545 Bk B In

-5.88

-5.51

-0.37 2

4 1803 545 ARO 0.77 1.49

-0.72 1563 545 Bk B In

-3.64

-3.95 0.31, 2

5 1620 545 ARO

-0.82

-0.49

-0.33 1208 545 Bk 6-1 In

-8.60

-9.39 0.79 3

1 823 545 ARO

-4.50

-4.35

-0.15 484 545 Bk 6-1 In

-15.12

-15.34 0.22 3

2 1175 545 ARO 0.52 2.34

-1.82 968 545 Bk B In

-5.70

-4.88

-0.82 3

3 1550 545 ARO 0.43 1.57

-1.14 1369 545 Bk B In

-6.13

-5.75

-0.38 3

4 1822 545 ARO 1.13 1.81

-0.68 1403 545 Bk 6-1 In

-6.60

-6.80 0.20 These HZP results are combined with the HIFP results in the following Table 4.10 for a total of 41 measurements for the statistical analysis.

36

SCE-0901 TABLE 4.10 At-Power Isothermal Temperature Coefficient - Cycles 1 - 5 Unit Cycle Power Burnup CBC

(%)

(GWD/T)

(PPM)

MEAS CASMO-4 M-C4 ITC ITC ITC (pcm/OF)

(pcm/IF)

(pcm/F) 1 20 0.103 50 0.539 80 1.250 100 2.050 100 9.180 2

2 98 0.208 100 1.466 100 6.650

100, 8.123 2

3 100 0.380 100 1.336 100 10.202 100 12.762 2

5 100 1.464 3

1 50 0.288 98 1.360 100 9.067 3

2 50 0.150 89 0.378 3

3 100 1.447 100 9.867 3

4 100 1.520 Bias from 41 measurements = -0.46 pcm / 'F 660 559 512 483 287 818 693 268 145 818 693 268 145 1063 540 471 277 893 758 991 367 1255

-6.28

-8.24

-9.42

-10.37

-16.47

-7.30

-12.50

-22.30

-25.42

-7.81

-9.23

-19.20

-23.00

-9.83

-8.26

-10.72

-14.78

-5.59

-10.84

-9.64

-22.20

-8.23

-6.39

-8.13

-9.43

-10.75

-15.16

-7.41

-10.11

-19.81

-22.69

-7.54

-8.66

-21.05

-25.15

-10.55

-8.89

-11.31

-15.11

-3.62

-7.80

-9.19

-20.72

-7.38 0.11

-0.11 0.01 0.38

-1.31 0.11

-2.39

-2.49

-2.73

-0.27.

-0.57 1.85 2.15.

0.72 0.63:

0.59' 0.33"

-1.97

-3.04

-0.45

-1.48

-0.85 CASMO-4 Uncertainty -- Normality Test Result = Normal Sample Size = 41 Degrees Of Freedom = 40 K 95/95 = 2.118 Standard Deviation (S) = 1:18 pcm/ 0 F 95/95 Uncertainty (K95/95

• S) = 2.50 pcm T OF 37

SCE-0901 TABLE 4.11 Power Coefficients SONGS Units 2&3 - Cycles 1, 2, And 3 Measured Pow Coeff (pcm/%AP)

CASMO-4 Pow Coeff (pcm/%AP)

(M - C4)

(pcm/%AP)

Unit Cycle Power Burnup Boron MWD/T Ppm 2

2 2

2 2

3 3

1 1

1 2

3 1

1 50 80 100 98 100 50 100 539 1250 2050 208 380 288 1360 559 512 483 818 1095 540 471

-11.04

-9.46

-9.47

-9.90

-11.03

-10.41

-8.93

-10.06

-9.61

-9.06

-9.38

-9.01

-10.04

-9.25

-0.98 0.15

-0.41

-0.52

-2.02

-0.37 0.32'

-0.55 0.78 Bias =

Standard Deviation =

38

SCE-0901 TABLE 4.12 Control Rod Bank Worths - Unit 2 BOC, HZP, Measured By Boration/Dilution Unit Cycle CEA Measured C-4/S-3 Predicted 100*(M-P) / P Group Worth (pcm)

Worth (pcm)

(%)

2 10 5

316 334

-5.4 4

615 667

-7.9 3

752 720 4.4 2

367 369

-0.6 2

11 5

385 380 1.2 4

634 641

-1.1 3

681 639 6.6 2

389 372 4.5 2

12 5

394 381 3.3 4

759 754 0.7 3

863 808 6.7 1

551 559

-1.5 6

401 413

-3.0 A

1369 1333 2.7 2

13 5

305 298 2.4 4

626 651

-3.9 3

893 841 6.2 2

409 381 7.2 6

359 372

-3.4 B

1812 1798 0.8 2

14 5

356 340 4.7 4

709 727

-2.5 3

786 717 9.6 1

708 711

-0.4 6

385 403

-4.6 A

1414 1356 4.3 39

SCE-0901 TABLE 4.13 Control Rod Bank Worths - Unit 3 BOC, HZP, Measured By Boration/Dilution Unit Cycle CEA Measured C-4/S-3 Predicted 100*(M-P) / P Group Worth (pcm)

Worth (pcm)

(%)

3 10 5

401 407

-1.4 4

640 703

-8.9 3

786 762 3.2 2

393 377 4.1 3

11 5

399 390 2.4 4

686 708

-3.1 3

815 761 7.1 2

378 362 4.3 3

12 5

317 306 3.6 4

712 734

-3.0 3

863 777 11.1 1

527 562

-6.2 6

386 399

-3.4 A

1323 1291 2.5 3

13 5

337 327 3.0 4

684 687

-0.4 3

812 738 10.0 2

434 396 9.7 6

369 371

-0.4 B

1698 1675 1.3 3

14 5

311 307 1.3 4

670 687

• -2.4 3

824 781 5.5 1

538 550

-2.2 6

383 397

-3.6 A

1451 1403 3.4 For All Unit 2 and Unit 3 CEA Bank Worth Measurements:

CASMO-4 Uncertainty -- Normality Test Result = Normal Sample Size = 52 Degrees Of Freedom = 51 K95/95 = 2.055 Bias (Mean) = 1.3%

Standard Deviation (S) = 4.67 %

Measurement Uncty: SM2 = 8.00%

Sc = [ (4.67)2 - 8.0 ]112 = 3.72%

95/95 Uncertainty (K95/95

• Sc) = 7.6 %

40

SCE-0901 TABLE 4.14 Net (N-i) Control Rod Worth CASMO-3 And CASMO-4 Comparison For Unit 2 Cycle 14 CASMO-3 CASMO-3 k-eff Worth (pcm)

CASMO-4 CASMO-4 k-eff Worth (pcm)

ARO ARI Rod 1 Rod 2 Rod 3 Rod 4 Rod 5 Rod 6 Rod 7 Rod 8 Rod 9 Rod 10 Rod 11 Rod 12 Rod 13 Rod 14 Rod 15 Rod 16 Total Worst Stuck (N-1) 0.99780 0.92439 0.92550 0.92656 0.92658 0.92584 0.92489 0.92613 0.92680 0.92686 0.92573 0.92656 0.92765 0.92794 0.92799 0.92637 0.92687 0.92529 Worth =

worth =

Worth =

130 253 256 169 58 203 281 288 157 253 380 414 420 231 289 105 7959 420 7539 1.00000 0.92543 0.92659 0.92770 0.92770 0.92693 0.92586 0.92724 0.92793 0.92794 0.92665 0.92769 0.92858 0.92909 0.92895 0.92746 0.92773 0.92631 135 264 264 175 50 211 291 292 142 263 367 426 409 237 268 103 8058 426 7632

% Difference

-1.2

-1.4

-1.2 41

SCE-0901 TABLE 4.15 Inverse Boron Worth BOC, Hot Zero Power MEAS IBW Unit Cycle 2

2 2

2 2

3 3

3 3

10 11 12 13 14 10 11 13 14 HZP HZP HZP HZP HZP HZP HZP HZP HZP (ppm/% Rho) 135.6 129.0 138.7 149.0 138.7 139.8 127.3 136.0 132.0 CASMO4 IBW (ppm/%Rho) 140.1 137.2 141.0 142.5 143.1 141.4 142.5 143.7 144.9 M-C4 IBW (PCT Diff)

-3.2

-6.0

-1.7 4.6

-3.0

-1.2

-10.6

-5.3

-8.9 Bias =

-3.9 CASMO-4 Uncertainty -- Normality Test Result = Normal Sample Size = 9 Degrees Of Freedom = 8 K95/95 = 3.031 Standard Deviation (S) = 4.5 %

95/95 Uncertainty (K95/95

• S) = 13.6 %

42

SCE-0901 SECTION 5 POWER DISTRIBUTION AND PIN PEAKING COMPARISONS

5.0 INTRODUCTION

The ability of CASMO-4 / SIM-ULATE-3 (C-4/S-3) to accurately calculate core radial and axial power distributions and pin peaking factors (Fq, Fxy, Fr) is verified by comparison with CASMO-3 / SIMULATE-3 (C-3/S-3) results.

If C-4/S-3 and C-3/S-3 core radial and axial power distributions, and pin peaking factors (Fq, Fxy, Fr) are essentially the same, the C-3/S-3 assembly and fuel pin peaking biases and uncertainties from Table 1.2 of Reference 1 also apply to C-4/S-3.

Section 5.1 compares local pin power.

Section 5.2 compares pin peaking factors (Fq, Fxy, Fr).

Section 5.3 compares core radial and axial power distributions.

Section 5.4 compares the axial shape index.

5.1 LOCAL PIN POWER The local pin power bias and uncertainty is determined by modeling industry standard critical experiments in which the individual fuel rod power distribution has been measured.

This bias and uncertainty were determined by analysis of the RPI critical experiments (Reference 2) with and without erbia and B&W critical experiments (Reference 2) with and without gadolinia.

The RPI core configurations calculated with CASMO-4 / DOT had 0, 20, 44, and 56 erbia pins.

A mean and standard deviation were calculated for each core configuration, and for the pooled Data. The mean was negligibly small and negative. Therefore it was assumed to be 0.

From Reference 2, the Local Pin Power bias and uncertainty to be used for SONGS 2 and 3 are:

0%

+ 1.78%

43

SCE-0901 5.2 PIN PEAKING FACTORS (Fq, Fxy, Fr)

This section compares CASMO-4/SEVIULATE-3 (C-4/S-3) calculated pin peaking factors (Fq, Fxy, and Fr) with CASMO-3/SIMJULATE-3 (C-3/S-3) calculated results.

SONGS 2&3 Cycles 12, 13, and 14 pin peaking factor comparisons are shown in Tables 5.1, 5.2, and 5.3. The C-3/S-3 and C-4/S-3 pin peaking factors are all less than about 1.5%.

Since the pin peaking factors are essentially the same, the C-3/S-3 biases and uncertainties also apply to C-41S-3.

Therefore the C-3/S-3 fuel pin peaking biases and uncertainties from Table 1.2 of Reference 1 also apply to C-4/S-3:

Bias

+

Uncertainty Fuel Rod Fq 0.0%

+ 4.62%

Fuel Rod Fxy 0.0 %

+ 5.20 %

Fuel RodFr 0.0%

+ 3.89%

These pin peaking factor comparisons demonstrate the continuing accuracy of SCE's neutronics models and methodology.

44

SCE-0901 5.3 CORE RADIAL AND AXIAL POWER DISTRIBUTIONS This section compares CASMO-4/STMIJLATE-3 (C-4/S-3) calculated radial and axial power distributions with CASMO-3/SEMULATE-3 (C-3/S-3) calculated results.

SONGS 2 Cycle 14 BOC and EOC radial and axial power distributions are shown in Figures 5.1, 5.2, 5.3, and 5.4. The C-3/S-3 and C-4/S-3 power distributions are negligibly different.

Since the power distributions are essentially the same, the C-3/S-3 biases and uncertainties also apply to C-4/S-3.

Therefore the C-3/S-3 assembly biases and uncertainties from Table 1.2 of Reference 1 also apply to C-4/S-3:

Bias

+

Uncertainty AssemblyFq 0.0%

+

4.17%

Assembly Fxy 0.0 % +

4.80 %

Assembly Fr 0.0%

+

3.34%

These power distribution comparisons demonstrate the continuing accuracy of SCE's neutronics models and methodology.

5.4 AXIAL SHAPE INDEX Axial Shape Index (ASI) is defined as:

(PB - PT) / (PT + PB) where PT = Power in the top half or the reactor core PB = Power in the bottom half or the reactor core CASMO-3 / STMULATE-3 and CASMO-4 / SIMULATE-3 BOC and EOC axial power shapes are compared in Figures 5.3 and 5.4. The axial shapes from the two computer program systems are negligibly different..

The CASMO-3 and CASMO-4 Axial Shape Index was compared for Cycle 14 of Units 2 and 3.

The average difference is 0.002, which is much smaller than the CASMO-3 Axial Shape Index uncertainty of 0.014.

Therefore the CASMO-3 Axial Shape Index bias and uncertainty apply to CASMO-4:

0.003 + 0.014 45

SCE-0901 TABLE 5.1 Power Distribution Peaking Factor -- Fxy 100% Hot Full power, All Rods Out Unit Cycle Burnup C-3/S-3 C-4/S-3 Difference Difference (GWD/MTU)

Fxy Prediction Fxy Prediction (C4 - C3)*

(%)**

2 12 0.25 1.435 1.430

-0.005

-0.35 1

1.424 1.421

-0.003

-0.21 2

1.420 1.416

-0.004

-0.28 4

1.421 1.416

-0.005

-0.35 6

1.414 1.409

-0.005

-0.35 8

1.402 1.401

-0.001

-0.07 10 1.393 1.399 0.006 0.43 12 1.379 1.392 0.013 0.94 14 1.363 1.379 0.016 1.17 16 1.349 1.364 0.015 1.11 18 1.336 1.347 0.011 0.82 20 1.320 1.331 0.011 0.83 2

13 0.25 1.398 1.407 0.009 0.64 1

1.380 1.390 0.010 0.72 2

1.368 1.377 0.009 0.66 4

1.374 1.369

-0.005

-0M36 6

1.375 1.370

-0.005

-0.36 8

1.368 1.373 0.005 0*37 10 1.360 1.375 0.015 1.10 12 1.354 1.369 0.015 1.11 14 1.344 1.360 0.016 1.19 16 1.336 1.350 0.014 1.05 18 1.327 1.338 0.011 0.83 20 1.312 1.324 0.012 0.91 2

14 0.25 1.414 1.430 0.016 1.13 1

1.397 1.412 0.015 1.07 2

1.388 1.398 0.010 0.72 4

1.396 1.391

-0.005

-0.36 6

1.391 1.387

-0.004

-0.29 8

1.380 1.387 0.007 0.51 10 1.367 1.385 0.018 1.32 12 1.359 1.377 0.018 1.32 14 1.351 1.366 0.015 1.11 16 1.341 1.356 0.015 1.12

.18 1.329 1.343 0.014 1.05 20 1.316 1.328 0.012 0.91

• CASMO4 - CASMO3

• • 100% x (CASMO4 - CASMO3)/(CASMO3) 46

SCE-0901 TABLE 5.2 Power Distribution Peaking Factor -- Fr 100% Hot Full power, All Rods Out Unit Cycle 2

12 2

13 Burnup (GWD/MTU) 0.25 1

2 4

6 8

10 12 14 16 18 20 0.25 1

2 4

6 8

10 12 14 16 18 20 0.25 1

2 4

6 8

10 12 14 16 18 20 C-3/S-3 Fr Prediction 1.402 1.403 1.408 1.405 1.399 1.389 1.377 1.366 1.353 1.340 1.327 1.313 1.352 1.358 1.365 1.371 1.370 1.364 1.354 1.343 1.332 1.319 1.309 1.298 1.371 1.378 1.387 1.391 1.385 1.375 1.362 1.348 1.335 1.323 1.314 1.302 C-4/S-3 Difference Difference Fr Prediction (C4 - C3)*

(%)**

1.399 1.401 1.405 1.404 1.398 1.389 1.384 1.376 1.365 1.352 1.338 1.322 1.352 1.354 1.361 1.367 1.369 1.365 1.364 1.358 1.349 1.336 1.323 1.312 1.370 1.374 1.383 1.389 1.385 1.378 1.374 1.365 1.351 1.338 1.328 1.316

-0.003

-0.002

-0.003

-0.001

-0.001 0.000 0.007 0.010 0.012 0.012 0.011 0.009 0.000

-0.004

-0.004

-0.004

-0.001 0.001 0.010 0.015 0.017 0.017 0.014 0.014

-0.001

-0.004

-0.004

-0.002 0.000 0.003 0.012 0.017 0.016 0.015 0.014 0.014

-0.21

-0.14

-0.21

-0.07

-0.07 0.00 0.51 0.73 0.89 0.90 0.83 0.69 0.00

-0.29

-0.29

-0.29

-0.07 0.07 0.74 1.12 1.28 1.29 1.07 1.08

-0.07

-0.29

-0.29

-0.14 0.00 0.22 0.88 1.26 1.20 1.13 1.07 1.08 2

14

• CASMO4 - CASMO3

• • 100% x (CASMO4 - CASMO3)/(CASMO3) 47

SCE-0901 TABLE 5.3 Power Distribution Peaking Factor -- Fq 100% Hot Full power, All Rods Out Unit Cycle Burnup C-3/S-3 C-4/S-3 Difference Difference (GWD/MTU)

Fq Prediction Fq Prediction (C4 - C3)*

(%)**

2 12 0.25 1.640 1.639

-0.001

-0.06 1

1.626 1.628 0.002 0.12 2

1.619 1.621 0.002 0.12 4

1.569 1.576 0.007 0.45 6

1.530 1.535 0.005 0.33 8

1.515 1.518 0.003 0.20 10 1.506 1.516 0.010 0.66 12 1.499 1.515 0.016 1.07 14 1.484 1.503 0.019 1.28 16 1.470 1.488 0.018 1.22 18 1.461 1.475 0.014 0.96 20 1.447 1.459 0.012

.0.83 2

13 0.25 1.571 1.586 0.015 0.95 1

1.555 1.560 0.005 0.32 2

1.553 1.553 0.000 0.00 4

1.516 1.519 0.003 0.20 6

1.482 1.486 0.004 0.27 8

1.475 1.485 0.010 0.68 10 1.471 1.488 0.017 1.16 12 1.471 1.490 0.019 1.29 14 1.462 1.481 0.019 1.30 16 1.455 1.470 0.015 1.03 18 1.449 1.463 0.014 0.97 20 1.436 1.450 0.014 0.97 2

14 0.25 1.593 1.615 0.022 1.38 1

1.587 1.589 0.002 0.13 2

1.587 1.586

-0.001

-0.06 4

1.545 1.550 0.005 0.32 6

1.502 1.508 0.006 0.40 8

1.487 1.500 0.013 0.87 10 1.478 1.500 0.022 1.49 12 1.477 1.498 0.021 1.42 14 1.470 1.488 0.018 1.22 16 1.460 1.477 0.017 1.16 18 1.452 1.469 0.017 1.17 20 1.439 1.455 0.016 1.11

• CASMO4 - CASMO4 100% x (CASMO4 - CASMO3)/(CASMO3) 48

SCE-0901 FIGURE 5.1 S2C14 Axially Integrated RPD Comparison 100% HIFP -- ARO -- BOC CASMO3 CASMO4 CAS4 - CAS3 0.381 0.754 0.377 0.748

-0.004

-0.006 Max Abs Diff =

0.005 Min Abs Diff =

-0.006 RMS Diff =

0.003 0.436 0.435

-0.001 0.327 0.326

-0.001 0.580 0.580 0.000 0.920 0.916

-0.004 0.381 0.994 0.377 0.990

-0.004

-0.004 0.754 0.852 0.748 0.852

-0.006 0.000 0.441 0.881 0.441 0.886 0.000 0.005 1.067 1.209 1.071 1.211 0.004 0.002 0.988 1.263 0.993 1.263 0.005 0.000 1.197 1.066 1.196 1.069

-0.001 0.003 1.030 1.262 1.032 1.261 0.002

-0.001 1.093 1.070 1.097 1.072 0.004 0.002 0.324 0.579 0.919 0.996 0.852 0.324 0.578 0.915 0.992 0.852 0.000

-0.001

-0.004

-0.004 0.000 1.058 0.985 1.196

.1.031 1.093 1.061 0.989 1.195 1.034 1.097 0.003 0.004

-0.001 0.003 0.004 1.200 1.259 1.065 1.263 1.070 1.201 1.259 1.068 1.261 1T.072, 0.001 0.000 0.003

-0.002 0.002 1.038 1.083 1.284 1.089 1,.286 1.041 1.086 1.283 1.090 1.284 0.003 0.003

-0.001 0.001

-0.002 1.084 1.282 1.101 1.287 1.088 1.087 1.281 1.103 1.284 1.089 0.003

-0.001 0.002

-0.003 0.001 1.285 1.101 1.285 1.065 1.249 1.283 1.103 1.283 1.067 1.246

-0.002 0.002

-0.002 0.002

-0.003 1.089 1.287 1.065 1.219 0.968 1.090 1.285 1.067 1.217 0.969 0.001

-0.002 0.002

-0.002 0.001 1.286 1.088 1.249 0.968 0.789 1.284 1.089 1.246 0.969 0.785

-0.002 0.001

-0.003 0.001

-0.004 49

SCE-0901 FIGURE 5.2 S2C14 Axially Integrated RPD Comparison 100% -FP -- ARO -- EOC CASMO3 CASMO4 CAS4 - CAS3 0.494 0.482

-0.012 Max Abs Diff -

0.006 Min Abs Diff =

-0.012 RMS Diff =

0.005 0.505 0.497

-0.008 0.391 1.015 0.384 1.012

-0.007

-0.003 0.662 0.975 0.654 0.975

-0.008 0.000 1.022 1.231 1.015 1.231

-0.007 0.000 0.494 1.132 1.019 0.482 1.130 1.020

-0.012

-0.002 0.001 0.843 0.913 1.049 0.834 0.912 1.052

-0.009

-0.001 0.003 0.503 0.494

-0.009 0.894 0.892

-0.002.

1.212 1.211

-0.001 1.254 1.255 0.001 1.018 1.021 0.003 1.244 1.248 0.004 1.003 1.007 0.004 0.390 0.662 1.023 1.133 0.384 0.654 1.016 1.130

-0.006

-0.008

-0.007

-0.003 1.013 0.975 1.231 1.019 1.010 0.974 1.231 1.021

-0.003

-0.001 0.000 0.002 1.209 1.253 1.018 1.245 1.209 1.255 1.021 1.248 0.000 0.002 0.003 0.003 0.993 1.016 1.248 1.014 0.994 1.020 1.252 1.018 0.001 0.004 0.004 0.004 1.016 1.237 1.016 1.242 1.020 1.241 1.021.

1.247 0.004 0.004 0.005 0.005 1.248 1.016 1.239 0.997 1.252 1.021 1.245 1.002 0.004 0.005 0.006 0.005 1.014 1.242 0.997 1.201 1.018 1.247 1.002 1.207 0.004 0.005 0.005 0.006 1.245 1.009 1.227 0.942 1.249 1.014 1.232.

0.947 0.004 0.005 0.005 0.005 0.844 0.835

-0.009 0.913 0.912

-0.001 1.049 1.052 0.003 1.003 1.007 0.004 1.245 1.249 0.004 1.009 1.014 0.005 1.227 1.232 0.005 0.942 0.947 0.005 0.811 0.815 0.004 50.

SCE-0901 FIGURE 5.3 S2C14 Average Axial Power Distribution Comparison 100% HEFP -- ARO -- BOC Axial Power Comparison 1.20 1.10-1.00 L, 0.90 0.80 0.70 0.60--

0.50 0

5 10 15 Axial Node --- CASMO-3 -A--- CASMO-4 20 51

SCE-0901 FIGURE 5.4 S2C14 Average Axial Power Distribution Comparison 100% HFP -- ARO -- EOC Axial Power Comparison 1.20-1.10 -

LL. 1.00-0.90 0.80 -

0.70 0

5

.10 15 20 Axial Node CASMO-3 --A-CASMO-4 52

SCE-0901 SECTION 6 FUTURE BURNABLE ABSORBERS AND FUEL ROD CLADDING

6.0 INTRODUCTION

The CASMO-4 / SIMULATE-3 methodology has also been extensively used for two other widely utilized, common burnable absorber types:

1) Gadolinia where (Gd20 3) burnable absorber is mixed directly with the U0 2 in the fuel rod,

2) Integral Fuel Burnable Absorbers (IFBA) where zirconium diboride (ZrB2) burnable absorber is applied in a thin layer directly on the surface of the U0 2 fuel pellet.

The use of these two modern burnable absorber types permits many more options for cycle reactivity management. Of the two types of burnable absorbers, gadolinia is far more challenging to model properly (IFBA with B-10 is fairly benign). CASMO-4 with its internal Gd-depletion capability was specifically designed to model gadolinia bearing fuel where resonance self-shielding is extremely important along with properly modeling the spatially heterogeneous

("onion-like") depletion of gadolinia.

References 12 and 20 provide benchmark results against critical experiments for gadolinia and MCNP comparisons for both gadolinia and TEBA burnable absorbers. References 18 and 19 provide further benchmarking of the CASMO-4 gadolinia depletion model.

6.1 CRITICALS WITH GADOLINIA Reference 12 documents the benchmarking of CASMO-4 against several sets of critical experiments. These sets include the Kritz critical experiments performed at Studsvik in Sweden on PWR 3x3 and BWR 4x4 assembly configurations, and the B&W critical experiments performed using PWR -5x5 assembly mock-ups.

The Kritz series of critical experiments are of special interest because many of the criticals were performed "at temperature," e.g., not just at cold conditions.

KRITZ-3 In the Kritz-3 PWR series or criticals four cores were analyzed - two U02 cores and two MOX cores - where measurements were performed at temperatures ranging from 20 C to 245 C (i.e.,

cold conditions to hot-zero-power conditions). The cores are relatively small, but at the same time exhibit low radial leakage.

53

SCE-0901 Summary of Results from Kritz-3 PWR Analysis Core Design k-eff 1 sigma U02 Cold Criticals 0.99874 0.00130 U02 Hot Criticals 0.99911 0.00127 MOX Cold Criticals 1.00055 0.00110 MOX Hot Criticals 0.99960 0.00075 The documented uncertainty in the measured boron concentrations was - 1% which is substantial because the boron concentrations were typically -1000 ppm which translates into -200 pcm in reactivity. From this, the CASMO-4 results may be considered to be well within the uncertainty of the measurements.

Although the KRITZ-3 PWR cores did not contain gadolinia (or IFBA), comparisons to this set of criticals provides a good estimate of the accuracy of the CASMO-4 isothermal temperature coefficient calculation.

KRITZ-4 The Kritz-4 series of criticals, do contain gadolinia and are of direct relevance to CASMO-4 benchmarking for this type of burnable absorber.

In the Kritz-4 series of criticals, two separate core configurations were modeled at both cold and hot-zero-power operating conditions again providing a good estimate of the CASMO-4 isothermal temperature calculation. The configurations below represent a variety of cores some with Gd (of various amounts) and some with control rods.

Core Condition Gd k-eff

% RMS'

%tfRMS 2

%1RMS3 2:1 Cold No 1.00060 0.89 0.73 1.02 2:2 Cold! Rodded No 1.00103 2:3.

Cold Yes 1.00079 2:3 Hot Yes 1.00021 0.90 0.63 1.33 2:4 Cold / Rodded Yes 1.00060 2:5 Cold Yes 1.00103 3:1 Cold No 1.00132 3:1 Hot No 1.00008 1.17 0.73 1.38 3:2 Cold Yes 1.00152 3:2 Hot Yes 0.99949 1.14 1.07 1.18 54

SCE-0901 3:3 Cold Yes 1.00256 3:3 Hot Yes 1.00006 1.02 1.04 1.45 3:4 Cold Yes 1.00279 3:4 Hot Yes 1.00031 1.09 0.66 1.32 3:5 Cold Yes 1.00182 3:5 Hot Yes 1.00017 1.29 0.86 1.32 RMS' Assy (2,2)

RMS2 Assy (4,2)

RMS3 Assy (1,1)

Summary of Results from KRITZ-4Analysis Core Design k-eff 1 sigma Core 2 Cold Criticals 1.00081 0.00019 Core 2 Hot Criticals 1.00021 N/A Core 3 Cold Criticals 1.00184 0.00063 Core 3 Hot Criticals 0.99985 0.00046 From these criticals it can be seen that CASMO-4 does a very good job at predicting not only overall global reactivity level, but Gd-worth as well (which is extremely sensitive)..

B&W 1810 Series The B&W 1810 series set of critical experiments (Reference 12) was meant to represent realistic reactor configurations and consisted of -5x5 array of either.PWR 15x15 type assemblies (Babcock & Wilcox design) or PWR 14x14 type assemblies (Combustion Engineering design).

The central assembly was altered from one critical experiment to the next. Some cores contained gadolinia bearing fuel pins, some cores AgInCd or B4 C control rods, some cores hollow rods, etc. From this set of criticals, the following cores were analyzed with these results:

Description k-eff RMS' RMS2 I

Base B&W, no absorbers 1.00014 0.49 II 16 AgInCd rods in center assy 0.99986 V

28 Gd rods, 12 in center assy 0.99952 0.47 0.75 XII Split enrichment B&W, 1.00153 0.74 No absorbers XIII 16 B4C rods in center assy 1.00134 XIV 28 Gd rods, 12 in center assy 1.00069 0.83 1.26 55

SCE-0901 XVII Base CE, No absorbers 1.00208 0.71 XX 32 Gd rods, 16 in center assy 1.00151 0.83 0.81 1 Fission rate RMS (%) in central assembly 2 Fission rate RMS (%) along core diagonal OVERALL RESULTS FOR CRITICALS WITH CASMO-4 The overall reactivity, for all BWR and PWR cores reported in Reference 12, is 1.00032 with a standard deviation of 123 pcm. In these terms, two-thirds of all cores analyzed fall within an eigenvalue of 0.99900 and 1.00150. This encompasses hot and cold conditions for BWRs and PWRs, with both U0 2 and MOX fuel, with and without gadolinia, cruciform control rods, AgInCd and B4C control rods, borated and non-borated coolant, and circular and square geometries. In addition, all calculated fission rate RMS's were within the uncertainty of the measurements.

In the CASMO-3 Topical Report (Reference 1), CASMO-3 S]MAULATE-3 pin power distribution results are reported for three B&W cores (Table 5.1):

Case CASMO-3 RMS (%)

CASMO-4 RMS (%)

Core I 0.631 0.49 Core XII 0.872 0.74 Core XVIII 0.925 (Not Analyzed)

The CASMO-4 results (% RMS) are consistent with and generally better than the above CASMO-3 results.

56

SCE-0901 6.2 MCNP COMPARISONS FOR IFBA AND GADOLONIA LATTICES Reference 20 provides comparisons between MCNP and CASMO-4 for several IFBA configurations and several containing gadolinia. All MCNP calculations used 1 million histories, split into 50 batches of 20000 histories each. For each lattice design, the fission rate distribution calculated by CASMO-4 is compared against that calculated by MCNP. The accuracy of the fission rate calculation is emphasized since it reflects the accuracy of the overall calculational scheme used in CASMO-4.

WESTINGHOUSE 15x15 28 IFBA PINS Model: 15x15 Westinghouse design containing 28 IFBA pins (out of 204 fuel pins) with three different IFBA strengths: (i) lx (nominal loading), (ii) 3x (three times nominal loading), (iii) 5x (five times nominal loading).

Fission Rate RMS(%)

Case RMS(%)

No IFBA 0.424 28 IFBA (lx) 0.505 28 IFBA (3x) 0.374 28 IFBA (5x) 0.547 IFBA Worth (pcm)

Case MCNP lx

-3086 3x

-7989 5x

-11882 CASMO-4

% Difference

-3117 1.00

-7997 0.10

-11848

-0.29 No IFBA's Fission Rate Distribution (C47MCNP)/MCNP*100%

0.20 0.31 0.58

-0.66 0.42 0.75 0.63 0.42

-0.40

-0.46

-0.19 0.21 0.53

-0.40

-0.09

-0.37

-0.10 0.32

-0.55 0.00

-0.75 0.10 0.31 0.10 -0.20 -0.21 0.85 0.32 -0.21 -0.53 -0.53 28 IFBA's (lx) Fission Rate Distribution (C4-MCNP)/MCNP*100%

0.30 0.00 0.11 0.10 -1.38

-1.05 -0.39 0.00

-0.09 0.10

-0.75 0.19 0.00 0'.31 -0.58

-0.42 0.00 -0.31

-0.37 0.19 0.77 0.10 1.06

-0.28 1.03 0.73 0.59 0.22

-0.31 -0.31 -0.72 57

SCE-0901 28 IFBA's (3x) Fission Rate Distribution (C4-MCNP)/MCNP*100%

-0.10 0.36

-0.76

-0.56 0.36

-0.62 1.02 0.31 0.69

-0.10

-0.37

-0.09

-0.44 0.0

-0.40 0.00 0.31 0.30 0.55

-0.28 0.00 0.21

-0.19

-0.28

-0.30 0.10 0.29

-0.20 0.85 0.10 -0.50 28 IFBA's (5x) Fission Rate Distribution (C4-MCNP)/MCNP*100%

0.29 0.26

-0.75

-1.10 0.13 0.93

-0.20 0.20

-0.29

-0..18

-0.36 0.20

-0.10 1.28 0.47 0.00 0.10 0.20 0.46 0.66 0.00 0.82 0.74

-0.72

-0.10 0.40

-0.48 0.52

-0.89 -0.51 -1.47 Combustion Engineering 14x14 64 IFBA PINS Model: 14x14 Combustion Engineering design containing 64 IFBA pins (out of 192 fuel pins) with IFBA strength: (i) lx (nominal loading)

Fission Rate RMS(%)

Case RMS(%)

64 IFBA 0.408 64 IEFBA + B4C ROD 0.795 IFBA Worth (pcm)

Case 64 IFBA MCNP

-12014 CASMO-4

-11996

% Difference

-0.15 B4C Control Rod Worth (pcm) with 64 IFBA present Coefficient MCNP CASMO-4

% Difference B4C Rod Worth

-23208

-23035

-0.75 No IFBA's Fission Rate Distribution (C4-MCNP)/MCNP* 100% (800 ppm, no B4C rod)

-0.47

-0.20

-0.53

-0 43

-0.22 0.11

-1.05 0.41

-0.41

-0.51

-0.21 0.11

-0.38 0.09 0.19 0.78 -0.37

-0.10 0.68 0.92 0.47 0.19

-0.11 0.53 58

SCE-0901 64 IEBA's Fission Rate Distribution (C4-MCNP)/MCNP* 100% (800 ppm, no B4C rod)

-0.84 0.78 0.10 0.00 0.21 0.23

-0

.09 0.58 0.29

-0.30

-0.10 0.41 0.76 0.00

-0.67

-0.11 0.09 -0.09 0.50 -0.11 -0.30

-0.68

-0.44 0.00 64 IFBA's Fission Rate Distribution (C4-MCNP)/MCNP* 100% (800 ppm, B4C rod) 0.56 0.59 1.34 0.45 0.00 0.10 0.82 0.89 0.38

-1.34

-1.14 0.75 0.39

-0.72

-0.74

-0.23 -0.76 0.00 0.59 -0.78 -1;15

-1.85

-0.44 0.83 WESTINGHOUSE 15x15 16 Gd PINS

\\-

Model: 15x15 Westinghouse design containing 16 gadolinia pins (out of 204 fuel pins) with three different IFBA strengths: (i) Ix (nominal loading),- (ii) 3x (three times nominal loading),

(iii) 5x (five times nominal loading).

Gd Worth (pcm)

Case MCNP 16 Gd

-12874 CASMO-4

-13070

% Difference 1.52 No Gd Fission Rate Distribution (C4-MCNP)/IMCNP* 100% (800 ppm)

-0.19

-0.10

-0.67 0.47 0.51 0.42

-0.40

-0.20

-0.19

-0.36

-0.09

-0.49 0.72

-0.78

-0.28 0.00

-0.27 0.30

-0.62 0.18 0.50 0.00

-0.27 0.94 0.10

-0.41

-0.19 0.10

-0.21 0.11 1.29 16 Gd Fission Rate Distribution (C4-MCNP)/MCNP*100% (800 ppm) 0.90 0.71 0.41

-0.09 0.48 0.20 0.56 0.58 0.59

-0.70 0.09 0.30 0.71 0.19

-0.70

-0.52 0.83

-0.11

-0.45

-0.18 0.10

-0.30 0.'41

-0.47

-0.93 0.38 0.00 0.55

-0.28 0.00 0.55 59

SCE-0901 Combustion Engineering 14x14 16 Gd PINS Model: 14x14 Combustion Engineering design containing 16 Gd pins (out of 192 fuel pins)

Fission Rate RMS(%)

Case RMS(%)

No Gd 0.404 16 Gd 0.713 Gd Worth (pcm)

Case MCNP 16 Gd

-15777 CASMO-4

-15914

% Difference 0.87 No Gd Fission Rate Distribution (C4-MCNP)/MCNP*100% (800 ppm, no B4C rod) 0.27

-0.30

-0.21

-0.11 0.11 0.00 0.10

-0.10

-0.30

-0.31

-0.51 0.00

-0.29 0.35

-0.35 0.88 -0.17 0.61 0.69 0.00 0.29 0.57

-0.78

-0.49 16 Gd Fission Rate Distribution (C4-MCNP)/IMCNP*100% (800 ppm, no B4C rod)

-0.79

-0.28

-0.91

-0.11

-0.60

-0.85

-0.09 0.62

-0.52 0.49 0.31

-0.19 0.49

-1.07

-0.60 0.27 1.46 0.16 0.67 0.53 -0.18 0.33 1.69 0.09

6.3 CONCLUSION

FOR FUTURE BURNABLE ABSORBERS Therefore, based on the benchmarks presented above, CASMO-4 / SIMULATE-3 will also accurately model both gadolinia and IFBA based burnable absorbers in the San Onofre Units 2 and 3 reactor cores.

The above IFBA and gadolinia pin power distribution RMS differences are consistent with and generally better than the CASMO-3 results reported in Reference 1. Therefore, the CASMO-3 assembly and fuel rod biases and uncertainties are applicable to IFBA and gadolinia fuel designs.

60

SCE-0901 6.4 M5 CLADDING SONGS 2 and 3 fuel assemblies currently use Zircaloy-4 and Zirlo cladding. Lead Fuel Assemblies (LFAs) from AREVA will use M5 cladding (Reference 21). References 21 and 22 provide the chemical composition of Zircaloy-4, Zirlo, and M5 cladding materials.

Neutronically, these three cladding materials are insignificantly different and all may be modeled as Zircaloy-4 in CASMO-4.

The NRC safety evaluation (Reference 23) for ZIRLO states:

"The change in the cladding material from OPTIN to ZIRLO will have negligible effect on nuclear fuel performance since the primary change in physics properties is a small increase in neutron absorption attributable to the addition of niobium. An increase in neutron absorption of this magnitude has no effect on nuclear performance. Thus, no modifications were made to the nuclear engineering methodologies or computer codes. This is the same approach that was used for the previous application of ZIRLO. The staff agrees that the change would be negligible and, thus, finds this approach acceptable."

The same discussion would apply to M5 cladding.

Therefore, Zircaloy-4, ZIRLO and M5 cladding have similar neutronic properties and can all be modeled as Zircaloy-4 in CASMO-4.

61

SCE-0901 SECTION 7 CONCLUSION SCE has performed extensive benchmarking using the CASMO-4 / SIMULATE-3 methodology.

This effort consisted of comparisons of calculated physics parameters to SONGS measurements, comparison to another approved, benchmarked system (CASMO-3), and Studsvik-Scandpower critical experiment benchmarking and MCNP comparisons. The benchmarking includes erbia, gadolinia, and IFBA burnable absorbers. The CASMO-4 / SIMULATE-3 methodology also applies to Zircaloy-4, Zirlo, and M5 cladding. A set of biases and 95/95 (probability/confidence) tolerance limits for key physics parameters have been determined.

As CASMO-4 is the direct evolutionary descendent of CASMO-3 which SCE has previously licensed, the experience and proficiency that the SCE engineering staff has demonstrated with CASMO-3 is directly applicable to CASMO-4.

Based on the analyses and results contained in this report, SCE concludes that the CASMO-4 /

SIMULATE-3 methodology applies to steady-state SONGS reactor physics calculations. The accuracy of this methodology is sufficient for use in licensing applications, SONGS reload physics analyses, safety analyses inputs, startup predictions, core physics databooks, and, reactor protection system and monitoring system setpoint updates.

Based on the analyses and benchmark results contained in this report, SCE concludes that the CASMO-4 / SIMULATE-3 methodology may also be used for fuel assembly types containing gadolinia burnable absorber and Integral Fuel Burnable Absorber (IFBA). The biases and uncertainties listed in this report also apply to fuel assembly types containing gadolinia and IEFBA's.

This effort has also successfully demonstrated SCE's ability and proficiency to use the CASMO-4 / SIMULATE-3 computer program package.

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SCE-0901 SECTION 8 REFERENCES

1. NRC Letter To Southern California Edison Co, August 10, 1992,

SUBJECT:

ACCEPTANCE OF TOPICAL REPORT SCE-9001, "PWR REACTOR PHYSICS METHODOLOGY USING CASMO-3 / SIMULATE-3" FOR USE AT SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 (TAC NOS. M77846, M77843, AND M77844).

2. NRC Letter To Arizona Public Service Company, March 20, 2001,

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION (PVNGS), UNITS 1,2, AND 3-ISSUANCE OF AMENDMENTS ON CASMO-4/SIMULATE-3 (TAC NOS. MA9279, MA9280, AND MA9281).

3. NRC Letter To Nuclear Management Company LLC, September 13, 2000,

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 SAFETY EVALUATION ON TOPICAL REPORT, NSPNAD-8101, REVISION 2, "QUALIFICATION OF REACTOR PHYSICS METHODS FOR APPLICATION TO PRAIRIE ISLAND UNITS 1 AND 2" (TAC NOS. MA7997 AND MA7998).

4. NRC Letter To Virginia Electric And Power Company, March 12, 2003,

SUBJECT:

VIRGINIA ELECTRIC AND POWER COMPANY -ACCEPTANCE OF TOPICAL REPORT DOM-NAF-1,"QUALIFICATION OF THE STUDSVIK CORE MANAGEMENT SYSTEM REACTOR PHYSICS METHODS FOR APPLICATION TO NORTH ANNA AND SURRY POWER STATIONS" (TAC NOS. M1B5434, MB5436, AND MB5437).

5. NRC Letter To Omaha Public Power District, March 11, 2005,

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT (TAC NO. MC4304).

6. NRC Letter To Duke Energy Corporation, August 20, 2004,

SUBJECT:

FINAL SAFETY EVALUATION FOR DUKE TOPICAL REPORT DPC-NE-1005P, "NUCLEAR DESIGN METHODOLOGY USING CASMO-4 /

SIMULATE-3 MOX".

7. (A) "CASMO-4, A Fuel Assembly Burnup Program, User's Manual", SP-01/400, Revision 4, December 2004, Studsvik Scandpower Inc.

(B) "CASMO-4i A Fuel Assembly Burnup Program, Methodology",

STUDSVIK/SOA-95/2, 1995, Studsvik Of America Inc.

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8. "CMS-LINK, User's Manual", SOA-97/04, Revision 2, April 1999, Studsvik Scandpower Inc.

9. "'NTERPIN-3, Studsvik CMS Fuel Performance Code", SSP-01/430, Revision 0, June 2001, Studsvik Scandpower Inc.

10. "SIIMULATE-3, Advanced Three-Dimensional Two-Group Reactor Analysis Code",

SSP-95/15! Revision 3, July 2005, Studsvik Scandpower Inc.

11. Updated Final Safety Analysis Report San Onofre Nuclear Generating Station, Units 2 And 3.

12. SOA-94/13, "CASMO-4 Benchmark Against Critical Experiments", Studsvik of America, Inc.

13 CENPD-153-P-A, "INCA/CECOR Power Peaking Uncertainty", Combustion Engineering, Inc., Approved By NRC July 02, 1980.

14. ANSI N15.15-1974, American National Standard, "Assessment of the Assumption of Normality (Employing Individual Observed Values)".

15. SCR-607, "Factors For One-Sided Tolerance Limits And For Variables Sampling Plans", Sandia Corportion, March1963.

16. D. Knott, M. Edenius, "Validation of the CASMO-4 Transport Solution", Proceedings of Mathematical Methods and Supercomputing in Nuclear Applications, Vol. 2, p. 547 Karlsruhe, Germany, April 1993.

17. D. Knott, M. Edenius, "The Two-Dimensional Transport Solution Within CASMO-4",

Trans. Am. Nuc. So., Vol 68, p. 457, San Diego, CA, June, 1993.

18. Y. Kobayashi, E. Saji, A. Toba, "Gadolinia Depletion Analysis by CASMO-4", Trans.

Am. Nuc. So., Vol 69, p. 443, San Francisco, CA, Nov., 1993.

19. D. Knott, M. Edenius, "Comparisons of Gadolinium Depletion in CASMO-4 and CASMO-3", Trans. Am. Nuc. So., Vol 72, p. 367, Philadelphia, PA, June, 1995.

20. SOA-94/12, "CASMO-4 Benchmark Against MCNP", Studsvik of America, Inc.

21. BAW-10227P-A, "Evaluation Of Advanced Cladding And Structural Material (M5) In PWR Reactor Fuel", Framatome Cogema Fuels, Feb 2000.

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22. CENPD-404-P-A, "Implementation Of ZIRLO Cladding Material In CE Nuclear Power Fuel Assembly Designs", Westinghouse Electric Company LLC, November 2001.

23. NRC to Westinghouse Electric Company Letter, September 12, 2001,

SUBJECT:

Safety Evaluation Of Topical Report CENPD-404-P, Revision 0, Implementation Of ZIRLO Cladding Material In CE Nuclear Power Fuel Assembly Designs", (TAC NO. MB1035)

(ADAMS Accession Number ML012670041).

24. WCAP-16011-P-A, Revision 0, "Startup Test Activity Reduction Program",

Westinghouse Electric Company LLC, February 2005.

65