ML20207M754

Reload Design Methodology II

ML20207M754
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Site:	Oconee
Issue date:	10/31/1985
From:	DUKE POWER CO.
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References
DPC-NE-1002A, NUDOCS 8701130273
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i DUKE POWER COMPANY j OCONEE NUCLEAR STATION j RELOAD DESIGN METHODOLOGY 11  ;

DPC-NE-1002A .

OCTOBER 1985 1

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i OUKE POWER COMPANY OCONEE NUCLEAR STATION RELOAD DESIGN METHODOLOGY II J

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NUCLEAR REGULATORY COMMISSION SAFETY EVALUATION REPORT i

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UNITED STATES h

3j NUCLEAR REGULATORY COMMISSION

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/ October 1,1985 Dockets Nos. 50-269, 50-270 and 50-287 Mr. Hal B. Tucker Vice President - Nuclear Production Duke Power Company P. O. Box 33189 422 South Church Street Charlotte, North Carolina 28242

SUBJECT:

RELOAD DESIGN METHODOLOGY II REPORT Re: Oconee Nuclear Station, Units 1, 2 and 3 By letter dated April 3,1985 you requested review of the Technical Report DPC-NE-1002, "0conee Nuclear Station Reload Design Methodology II." We have reviewed the updated Oconee reload design methodology and the additional information submitted on June 12, 1985. The revised methodology is based on the CASMO computer code (instead of the EPRI-CELL / Color Set PDQ) but it is intended an an alternative rather than a replacement for the original methodology as described in NFS-1001A. Also, this report incorporates the analysis for minor changes in the fuel design, fuel mechanical performance and thermal-hydraulic design which resulted from the introduction of a new fuel assembly, the Mark B5-Z. The CASMO reload design methodology has been previously approved by us for the MCGuire and Catawba plants. The first use of this methodology will be for the Oconee Unit 1 Cycle 10 reload projected for February-March 1986.

We found the revised methodology to be an acceptable means for the calculation of the Oconee Station licensing submittals. The results of our review are contained in the enclosed Safety Evaluation.

Sincerely, hb.hw ,.

J hn F. Stolz, Chie Operating Reactors Branch #4 -

Division of Licensing

Enclosure:

Safety Evaluation I cc w/ enclosure:

See next page l

Mr. H. B. Tucker Oconee Nuclear Station Duke Power Company Units Nos. 1, 2 and 3 CC:

Mr. William L. Porter Duke Power Company P. O. Box 33189 422 South Church Street Charlotte, North Carolina 28242 J. Michael McGarry, III, Esq.

Bishop, Liberman, Cook, Purcell & Reynolds 1200 Seventeenth Street, N.W.

Washington, D.C. 20036 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Regional Administrator U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W.

Suite 3100 Atlanta, Georgia 30303 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control

-2600 Bull Street Columbia, South Carolina 29201 Office of Intergovernmental Relations -

116 West Jones Street Raleigh, North Carolina 27603 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621

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UNITED STATES Ei o NUCLEAR REGULATORY COMMISSION 3 .. (

• f. j WASHINGTON, D. C. 20555 Q~ck/ s,,,. -

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE RELOAD DESIGN METHODOLOGY II TECHNICAL REPORT CPC-NE-1002 FOR THE DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS N05. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287

1.0 INTRODUCTION

, The Duke Power Company Oconee Reload Design Methodology report NFS-1001A l was first approved by the NRC on July 29,1981(Ref.1). By letter dated April 3,1985 the Duke Power Company requested review of an alternative '

reload methodology which is described in technical report DPC-NE-1002 (Ref.2). This new methodology has been based on the CASMO code instead of the EPRI-CELL code that is described in NFS-1001A. The justification for Duke's request for an alternate methodology is that CASMO is less expensive to run yet just as accurate as EPRI-CELL. The CASMO code has been previous 1v aooroved by the staff for use in the analvsis of DFC's EcGuire and Catawba plants. Appropriate reliability factors have been developed for tne application of the CASMO code which is used to generate the physics data base. In addition DPC-NE-1002 deals with the updated fuel design, fuel mechanical performance and thermal-hydraulic design resulting from the introduction of the Mark B5-Z fuel assembly. The report covers the fuel design, fuel cycle' design, fuel mechanical and 3

thermal performance, maneuvering analysis, thermal-hydraulic design, -

Technical Specifications review and development, accident analysis and .

core physics parameter development. We shall review and evaluate each of ,

these sections.

2.0 EVALUATION 2.1 Fuel Design: The most important changes in the fuel design are the cladding thickness and the length cf tnc axial power shaping rods.

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However, these changes will be evaluated for each reload using the methodology described (Ref. 1).

2.2 Fuel Cycle Desion: The new methodology employs the CASMO code for the preparation of the nuclear cross sections for each material region in the fuel assembly; i.e., fuel pins, water holes, lumped burnable poisons, control rods and assembly gaps. The code prepares the table sets required for PDQ07 and also provides assembly average physics data required for the EPRI-N0DE-P calculations; therefore, no color set PDQ calculations are required. Once the PDQ and the EPRI-N0DE-P models have been developed, the fuel cycle design calculation proceeds as described in Section 3 of NFS-1001A (Ref. 1). In view of the favorable and acceptable comparisons of the calculations of this methodology to criticals as discussed in the report, this fuel design methodology is acceptable.

2.3 Fuel Mechanical and Thermal Performance: This section is essentially unchanged from the corresponding section of NFS-1001A. Each fuel reload is bounded by a reference design analyses; however, differences in fuel rod design, as fabricated dimensions or densification kinetics do occur.

Such changes must then be evaluated for: cladding creep collapse, cladding strain, cladding stress, fuel pin teg.perature, fuel pin pressure and ECCS analysis interface criteria.

Cladding Collapse. The methodology remains unchanged except that the approved TACO 2 (Ref. 3) is substituted for TACO. In addition the generic assembly power envelope has been revised to represent recent and current fuel cycle designs.

Cladding Strain Analysis. A generic strain analysis has been completed by Duke Power usin'g TAC 02 to ensure that the strain criterion of 1.0% total uniform strain is not exceeded. Should reanalysis be required due to significant changes in the fuel design, the reference analysis must be repeated using the TAC 02 code.

Duke Power Company ,

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Cladding Stress Analysis. The methodology of NFS-1001A has been revised to be consistent with that approved by the staff in the reload submittal for Unit 3, Cycle 8 and Unit 3 Cycle 9 (Refs. 4 and 5). The revision is in the ciasign stress intensity limits on mechanical properties which are based on the requirements of ASME code Article III-2000. The conservatisms of the original NFS-1001A cladding stress analysis are maintained and compliance with the ASME code criteria assures cladding integrity for the most limiting design conditions. This cladding stress analysis is acceptable.

Linear Heat Rate Capability. This methodology remains unchanged except that TACO 2 replaces TACO and results in an associated adjustment of the linear heat rates to reflect typical calculated values. The fuel rod axial power shape is assumed to be a 1.5 truncated cosine and the pin power envelope has been revised to reflect small recent changes in the pin power envelope vs burnup.

Finally a larger than nominal radial gap is employed which at. counts for manufacturing uncertainty distribution. The TAC 02 code has been approved by the staff and the refinements are conservative; hence, the linear heat rate analysis is acceptable.

Fuel Pin Pressure Analysis. This methodology remains unchanged except that TAC 02 replaced TACO and results in an associated adjuste m in the axial power shape and the generic pin power envelope vs burnup. The calculated interna 1 pressure values reflect typical values associated with TACO 2. Because TAC 02 has been approved by the staff and is used in a conservative manner, the .

fuel pin pressure methodology is acceptable.

ECCS Analysis Interface Criteria. The technical and design data for each fuel batch are reviewed by the Duke Power Company to assure that the license basis ECCS analyses are satisfied. However, if for a specific cycle the rod thermal analysis inputs lie out-Duke Power Company 3

side the reference analysis, then the licensee will reanalyze to ensure that the particular fuel batch meets the licensing criteria of 10 CFR 50.46. The responsibility for the identification of such incompatibility and its resolution lies with the licensee.

2.4 Maneuvering Analysis: The purpose of the maneuvering analysis is to generate three dimensional power distributions and potential imbalances for a variety of rod positions, xenon distributions and power levels. This analysis is not affected by the introduction of CASMO as the lattice code and Section 5 of. NFS-1001A (Ref.1) is applicable and acceptable.

2.5 Thermal-Hydraulic Design: The methodology of NFS-1001A remains unchanged

• cxcept that the introduction of CASMO changes the total nuclear uncertainty factor from 1.065 to 1.057. The thermal-hydraulic methodology in NFS-1001A is used to analyze the Mark-BZ reload assemblies. In NFS-1001A the CHATA (Ref. 6) and TEMP (Ref 7) codes are used for the thermal-hydraulic analysis of the core. However, using the BWC critical heat flux correlation (Ref. 8) in conjunction with CHATA (which has been approved by the NRC staff) the minimum DNBR design limit for normal operation and anticipated operational occurrences is 1.18. The rod bow penalty discussed in Section 6.10 of NFS-1001A is no longer applied. The DNBR penalty due to rod bow was estimated- to be insignificant because the power production capability of the fuel decreases with irradiation. A topical report (Ref. 9) just-ified this conclusion which has been accepted by the staff. The Mark-BZ assemblies use Zircaloy spacer grids; hence, the flux depression factor used to estimate the hot channel factor for the generic DNBR curves is now 1.007. The thermal-hydraulic analysis is based on methods approved for use by DPC and is acceptable. .

2.6 Technical Specifications Review and Development: The methodology for the process of Technical Specification development for the Oconee reloads remains the same as described in Section 7 of NFS-1001A (Ref.1) except for minor differences emanating from the adoption of the BWC CHF correlation Duke Power Company 4

and the CASMO code. The DNBR limit corresponding to a 95/95 probability /

confidence level for the BWC CHF correlation for the Mark-BZ fuel assembly' is 1.18. The uncertainties associated with the BWC CHF correlation have been reported in Reference 7 and have been approved by the staff. The setpoint calculations for the reactor protection system have been revised accordingly to ensure reactor protection system activation prior to exceeding any safety limits. The following Sections of NFS-1001A have been revised due to the use of CASMO.

Section 7.2.2.1 The nuclear uncertainty factor is 1.005, the engineering hot channel factor remains unchanged at 1.014. An 8% uncertainty is applied for all axial locations as an upper bound to account for fuel densification. All uncertainty factors are statistically combined to yield a centerline fuel melt uncertainty of 1.111.

Section 7.2.2.2 The methodology is unchanged by the use of CASMO but the radial uncertainty factor is 1.048.

Section 7.4.1 The methodology for the ECCS analysis power distribution remains unchanged. The effect on non-equilibrium xenon conditions is explicitly accounted for in the LOCA limits. The statistical combination of all the uncertainty components yields a total uncertainty factor of 1.070.

2.7 Accident Analysis Section 8 of NFS-1001A which discusses _the safety _ _ -

analysis methods in the Oconee reload design does not change due to the use of the CASM0 code. However, the average fuel temperature is revised to 2,950*F (frora 3,120*F) as a result of the Mark B5-Z fuel thermal performance.

Duke Power Company 5

2.8 Core Physics Parameter Develcoment The methods discussed in Section 9 of NFS-1001A which provide core physics data required for startup test predictions and the core physics report do not change due to the use of the CASMO code.

3.0 CONCLUSION

S We have reviewed the information presented in the revised reload design methodology for the Oconee Nuclear Station in licensing topical report OPC-NE-1002, including two supplements to the report and the additional information submitted on June 12, 1985 in response to a staff request.

This report parallels closely the original Oconee reload design methodology NFS-1001A. The main difference of the new methodology is the use of the CASM0 code for generating reactor physics data. In addition minor design changes are introduced in the fuel assembly design with the adoption of the Mark BS-Z element.

The report deals with each one of the corresponding chapters of NFS-1001A; i.e., fuel design, fuel cycle design, fuel mechanical performance, maneuvering analysis, thermal-hydraulic design, Technical Specification review and development, accident analysis and development of core physics parameters.

In each chapter the parts affected by the proposed changes were reanalyzed.

The new methodology report includes supplements which deal with methodology and nuclear reliability factors based on the CASMO lattice physics data.

The results of our review indicate that the CASMO based methodology can adequately predict soluble baron concentration, control rod worths, ejected rod worths and the thermal temperature coefficient. PDQ.and_CASM0_ radial -

peaking factors were compared and it was shown that CASMO overpredicted P0Q' results by a maximum of 1.4%.

Duke Power Company 6

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Although results from Unit 1 were used for this report the results and l conclusions are applicable to all three units because:

(a) all units are manufactured by the same vendor and use similar fuel, (b) all units have identical incore detector systems, and (c) calculations for all three units use the same methods and procedures.

We conclude that the proposed alternative reload design methodology is able to account for the main physics core parameters within the limits of reasonable uncertainty. Therefore, the proposed methodology in report DPC-NE-1002 is an acceptable alternative to the Oconee reload methodology presented in NFS-1001A and may be referenced by the Duke Power Company in licensing sub-mittals for the Oconee units.

Dated: October 1,1985 Principal Contributor: L. Lois Duke Power Company 7

References

1. NFS-1001A, " Duke Power Company, Oconee Nuclear Station, Reload Design Methodology", Duke Power Company, April 1984.

2. DPC-NE-1002, " Duke Power Company, Oconee Nuclear Station, Reload Design Methodology", Duke Power Company, March 1985.

3. Hsii, Y, et.al.. " TACO 2-Fuel Pin Performance Analysis", BAW-10141PA, Rev. 1, Babcock and Wilcox, Lynchburg VA, June 1983.

4. DPC-RD-2003, "0conee Unit 3, Cycle 8 Reload Report". Duke Power Company, February 1984.

5. DPC-RD-2005, "Oconee Unit 3, Cycle 9 Reload Report", Duke Power Company, May 1985.

6. BAW-10110, Rev.1, "CHATA-Core Hydraulics and Thermal Analysis", Babcock and Wilcox, Lynchburg VA, May 1977.

, 7. BAW-10021, " Thermal Energy Mixing Program, TEMP", Babcock and Wilcox, Lynchburg VA, April 1970.

8. Wilson, R. H. et. al., " Correlation of 15x15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation", BAW-10143PA Part 2, Babcock and Wilcox, Lynchburg VA, August 1981.

9. DPC-RD-2002, "Oconee Unit 2, Cycle 7 Reload Report", Duke Power Company September 1983.

Duke Power Company 8

DUKE POWER COMPANY TOPICAL REPORT

J 1

DUKE POWER COMPANY OCONEE NUCLEAR STATION RELOAD DESIGN METHODOLOGY II l

Technical Report t

DPC-NE-1002A 4

1 Octsbe , 1985 i

ABSTRACT This report revises certain aspects of Duke Power's Reload Design Methodology which have evolved since the NRC approval of NFS-1001A. Additionally, this report presents minor changes to the Fuel Design, Fuel Mechanical Performance, and Thermal-Hydraulic Design sections that are primarily the result of switching to the new fuel assembly, the Mark 85-2.

An alternative method for generating nuclear physics data is justified based on the CASMO computer code rather than the EPRI-CELL / color set PDQ sequence as described in NFS-1001A. A new set of reliability and uncertainty factors to be used for CASMO based reload designs is also presented. Other sections of this report have been modified to reflect the new reliability factors. It is emphasized that the CASMO methodology is presented as an alternative to the original methodology.

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TABLE OF CONTENTS .

P.aste 1- 1

1. INTRODUCTION

2. FUEL DESIGN 2-1

3. FUEL CYCLE DESIGN 3-1

4. FUEL MECHANICAL PERFORMANCE 4-1 4.1 Introduction 4-1 4.2 Cladding Collapse 4-1 4.3 Cladding Strain Analysis 4-2 4.4 Cladding Stress Analysis 4-3 4.5 Fuel Pin Pressure Analysis 4-4 4.6 Linear Heat Rate Capability 4-4 4.7 ECCS Analysis Interface Criteria 4-5

5. MANEUVERING ANALYSIS 5-1~

6. THERMAL-HYORAULIC DESIGN 6-1

7. TECHNICAL SPECIFICATIONS REVIEW AND DEVELOPMENT 7-1

8. ACCIDENT ANALYSIS REVIEW 8-1

9. DEVELOPMENT OF CORE PHYSICS PARAMETERS 9-1

10. REFERENCES 10-1 Appendix A - CODE

SUMMARY

Supplement 1 PHYSICS TEST COMPARISONS FOR THE CASMO METHODOLOGY Supplement 2 NUCLEAR RELIABILITY FACTORS FOR EPRI-N00E-P BASED ON CASMO LATTICE PHYSICS DATA iii l

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'2-1. 2-2 Oconee System and Component Data i

4-1 Fuel Mechanical Performance Assessment Criteria 4-6 7-1 Typical LOCA Kw/ft Limits for 0conee 7-5 s

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3-1 Nuclear Analysis Flow Chart for CASMO 3-2 4-1 Generic Pin Power Envelope 4-7 2 Generic Assembly Power Envelope 4-8

4-3 Pin Pressure History 4-9 4-4 Plot of LHRTM vs. Burnup 4-10 4

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1. INTRODUCTION Duke Power Company currently performs reload design analyses with the methodology of NFS-1001A (Reference 1). Thiis methodology is the NRC approved version with the revisions and safety evaluation report included. Since Reference 1 was approved, Duke Power has performed the reload design and licensing for Oconee Unit 2, Cycles 7 and 8 (References 2 and 3) and Oconee Unit 3, Cycles 7 and 8 (References 4 and 5).

Part of the Duke Power Reload Design effort involves reviewing and improving procedures, methods, and computer codes on a continuing basis. The development of Reload Design Methodology based on the CASMO-2 computer code is a result of this philosophy. This methodology uses CASMO-2 as a cross section generator and to determine assembly average physics data. The discussions of CASMO-2 in this report are applicable to all future versions and enhancements of CASMO which might be used by Duke Power Company. Hence, the lattice code is referred to simply as CASMO throughout this document.

This report justifies the use of CASMO for Oconee Reload Design calculations.

The change from EPRI-CELL to CASMO as the lattice code does not affect other computer models which are a part of the Reload Design process.

However, the change in the lattice code does produce subsequent changes in calculated core physics parameters (e.g. , power distributions, reactivity coef ficients, control rod worths, xenon worths, etc.) from the other models such as PDQ and EPRI-N00E-P.

1-1

CASMO is presented as an alternate lattice code to EPRI-CELL and appropriate reliability factors are developed to be used when CASMO generates the physics data base. Sections of the original methodology from NFS-1001A are modified in this report to reflect the appropriate uncertainties to be applied if CASMO methodology is used in place of EPRI-CELL.

Additionally, this report presents updates to the Fuel Design, Fuel Mechanical Performance, and The mal-Hydraulic Design Sections which have occurred since NFS-1001A was approved.

1-2

2. FUEL DESIGN There are no changes from Section 2 of Reference 1 except for numeric changes in Table 2-1, which are being included to reflect recent fuel assembly and

!. core component design changes.

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TABLE 2-1 OCONEE SYSTEM AND COMPONENT DATA MARK B5-Z CORE COMPONENT DESIGN DATA Fuel Assembly Designation Mark B5-Z Calculated Fuel Assembly Total Weight, Ibs. 1513 Weight by Material per Assembly, lbs.

Zircaloy 297.5 Inconel 9.7 UO 1160 2

Fuel Assembly Cell Flow Areas, in. 39.76 Assembly Pitch, in. 8.587 Spacer Grids End Spacer Grid Material Inconel 718 Intermediate Spacer Grid Material Zircaloy-4 Number of Spacer Grids per Assembly Total 8 Between Active Core Limits 6 Spacer Grid Weight, lbs.

Intermediate, each 1. 8 End, each 2.9 I

Total per Fuel Assembly 16.6 2-2

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Fuel Rods Fuel Rod Pitch, in. 0.568 Fuel Rod Array 15 x 15 Number of Fuel Rods per Assembly 208 l Fuel Rod Weight, lbs. s7.0 f.

I Fuel Clad Weight, lbs. 1.15 Fuel Rod Clad 00, in. 0.430 ,

Fuel Rod Clad 10, in. 0.377 Fuel Rod Clad Wall Thickness, in, nominal 0.0265 Fuel Rod Length, in. 153.6875 Fuel Rod Clad Material Zircaloy-4 l Fuel Rod Clad Material Density, lbs/in.3 0.237 Fuel Pellets Fuel Column Starts, From Bottom of Fuel Rods, in. 4-11/32 Fuel Pellet Diameter, in. 0.3686 Pellet Density, g/cc (95% of Theoretical) 10.41 Diametrical Gap, in. 0.0084 Fuel Column Length, in. 141.8 Unit Fuel Cell Flow Area,.in. 0.1774 Material 00 2

U per Fuel Ro'd, kg 2.229 U per Assembly, kg 463.6 UO2 per Fuel Rod, kg 2.528 U02 per Fuel Assembly, kg 525.9 U/UO Ratio used for Calculation 0.8815 2

2-3

Control Rod Guide Tubes No. of Control Rod Guide Tubes per Assembly 16 Guide Tube Material Zircaloy-4 Guide Tube Material Density, lbs/in.3 0.237 Weight of control rod guide tubes, lbs.

Per Control Rod Guide Tube 1.05 Total per Assembly 16.8 Guide Tube 00, in. 0.530 Guide Tube 10, in. 0.498 Guide Tube Wall Thickness, in. 0.016 Guide Tube Length, in. 156.3 Instrumentation Tube Number of Instrumentation Tubes per Assembly __

1 Instrumentation Tube Material Zircaloy-4 Instrumentation Tube Material Density, lbs/ft.3 0.237 Instrumentation Tube Weight, lbs.

Per Tube 1.40 Instrumentation Tube 00, in. 0.493 Instrumentation Tube 10, in. 0.441 Instrumentation Tube Wall Thickness, in. 0.026 Instrumentation Tube Cell Flow Area, in.2 0.0867 Length of Instrumentation Tube, in. 153.5 Spacer Sleeves Spacer Sleeve, 00, in. 0.550 Spacer Sleeve, 10, in. 0.498 2-4 i

' Spacer Sleeve Material Zircaloy-4 Spacer Sleeve Weight, 1bs.

Per Assembly 1.39 Control Rod.

Cladding 00, in. 0.441 Cladding ID, in. 0.398 Cladding Wall Thickness 0.023 Cladding Material Inconel-625 Absorber Length, in. 139.0 Absorber Material Ag-In-Cd Axial Power Shaping Rod Cladding 00, in. 0.440 Cladding ID, in. 0.386 Cladding Wall Thickness, in. 0.027 Cladding Material 55304 Absorber Length, in. 63 Absorber Material Inconel-600 Burnable Poison Ro_d Cladding 00, in. 0.430 Cladding 10, in. 0.360 Cladding Wall Thickness, in. 0.035 Cladding Material Zircaloy-4 Burnable Poison Length, in. 126.000 Burnable Poison Material A123 0 -84C 2-5

3. FUEL CYCLE DESIGN The development of the reactor core models used to perform the Fuel Cycle Design calculations is affected by the lattice code selected to generate the basic physics data. Reference 1 describes the calculational system using EPRI-CELL and the CASMO methodology is described below. Once the P0Q and EPRI-NODE-P core models are developed, the calculations and assessments performed to verify the acceptability of the preliminary and final fuel cycle designs are unchanged from Section 3 of Reference 1.

The nuclear calculational system with CASMO is summarized by Figure 3-1. The computer codes which are used are briefly described in the Appendix. CASMO is used to develop a set of nuclear cross sections which characterize each material region in the assembly including fuel pins, water holes, lumped burnable poisons, control rods, and assembly gaps.

In general, CASMO is used to model each unique combination of fuel enrichment and lumped burnable poison chosen for the fuel cycle design. The computer codes CHART and CASTHF are used to prepare the cross section tablesets required for P0Q07.

The same CASMO calculations which generate cross section data for the diffusion theory calculation also provide assembly average physics data required for EPRI-NODE-P nodal calculations. NORGE-P prepares the necessary EPRI-N00E-P input from the results of the CA3M0 calculations. Thus, no color set P0Q calculations are required with the CASMO methodology.

3-1

t Figure 3-1 NUCLEAR ANALYSIS FLOW CHART FOR CASMO CASMO A

NORGE-P CHART CASTHF V Y V EPRI-NODE-P < 2-0,1-CORE 4 P0Q07 with Thermal Feedback A

V EPRI-SHUFFLE V

3-0 Information 3-2

7..

4. FUEL MECHANICAL AND THERMAL PERFORMANCE 4.1 Introduction This section is unchanged from Reference 1 except for Table 4-1 and the first paragraph, wherein the " Vendor ECCS Analysis Interface Criteria" review is completed for each reload. The revised material is given below.

Each fuel cycle design requires that thorough fuel mechanical and thermal assessments be performed. A reload design utilizes fuel designs that are bounded by previous fuel assembly design analyses. However, differences in the-fuel rod design, as-fabricated dimensions, densification kinetics, or fuel cycle design do occur. These changes must then be assessed in regard to the following:

• Cladding creep collapse

• Cladding strain

• Cladding stress

• Fuel pin temperature

• Fuel pin pressure

• Vendor ECCS analysis interface criteria 4.2 Claddina Collapse The methodology remains unchanged except for the following:

1. TACO 2 (Reference 6) has been substituted for TACO.

2. Figure 4-2 has been revised to present a generic power history that is more representative of current fuel cycle designs.

4-1 i

4.3 Cladding Strain Analysis The limit on transient cladding strain is that uniform total strain of the cladding should not exceed 1.0%.

A generic strain analysis has been completed by Duke using TACO 2 to ensure that the strain criterion above is not exceeded. This is a bounding, generic analysis that requires no reload assessments.

Should reanalysis be required because of a significant change in the fuel rod design, Duke's generic strain analysis would be repeated using the same methodology. A description of the generic methodology follows:

TACO 2 is used to calculate cladding strain. A very conservative local power ramp is first determined by considering a maximum local power change induced by a worst case core maneuvering scenario. The scenario involves some or all of the following: core power level changes, xenon transient, ar.d control rod position changes. This worst case local power level change is then modeled in TACO 2 to determine the fuel pellet thermal expansion. The cladding transient strain is calculated from the pellet expansion using the following equation: l Strain = (Pellet 0.0.) peak - (Pellet 0.0.)o x 100 <1.0% __ ._

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(Pellet 0.0. ) .

where (Pellet 0.0.) peak = the maximum pellet 0.0. at the local power peak, and (Pellet 0.0.)g = pellet 0.0. prior to the ramp.

4-2

Pellet 0.0. dimensions are used to calculate cladding strain because the strain itself is caused by pellet thermal expansion. There are three major conditions in this calculation that make it conservative. The first is the extreme power change that is used to simulate the worst case peaking. The second is that the pellet is assumed to be in hard contact at inititation of the ramp. This is a conservative assumption since the power ramp is initiated from a very low power level and pellet / cladding contact is not expected to occur at this low linear heat rate. The third conservatism is that the pellet is non-compliant and that all of the pellet thermal expansion results directly in cladding strain.

4.4 Cladding Stress Analysis The methodology of Reference 1 has been revised. The current methodology is consistent with that of References 5 and 12.

The static stress analysis uses design stress intensity limits on mechanical

. properties based on the requirements of ASME Code Article III-2000. Thus, the design stress intensity value for Zircaloy-4 is the lowest of the following:

(1) one-third of the specified min.imum tensile strength at room temperature (2) one third of the tensile strength at temperature (3) two-thirds of the specified minimum yield strength at room temperature (4) two-thirds of the yield strength at temperature 4-3 1

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4.5 Fuel Pin pressure Analysis

-This methodology remains unchanged except for the following:

1. TACO 2 has replaced TACO

2. The fuel rod is assumed to have a 1.5 cosine axial flux shape at beginning of life and decreasing with burnup.

3. Figure 4-1 has been revised to indicate small changes in the " typical" generic pin power envelope.

4. Figure 4-3 has been revised to reflect the typical calculated values of internal rod pressure associated with the change to TAC 02.

4.6 Linear Heat Rate Capability This methodology remains unchanged except for the following:

1. TACO 2 has replaced TACO

2. The fuel rod is assumed to have a 1.5 cosine axial flux shape at beginning of life and decreasing thereafter.

3. Figure 4-1 has been revised to indicate small changes in the " typical" i generic pin power envelope.

4. Figure 4-4 has been revised to reflect the typical calculated values of linear heat rate capacility associated with the change to TACO 2.

5. A larger than nominal radial gap is employed which includes the effects of an LTL pellet 0.0. and UTL clad I.O. All statistics are performed at the 95/95 level (Reference 11).

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. ,-- -- , ,, ,,, - -----...-e --._ . . - - - . , - . - - . - . - -

4.7 ECCS Analysis Interface Criteria Duke reviews each batch of fuel and the fuel cycle design for compatibility with the vendor's fuel rod thermal analysis inputs to the ECCS analysis. This review considers all items indicated in Table 4-1. Review criteria have been developed by Duke and have been reviewed and approved by the vendor.

Should the fuel rod thermal analysis inputs for a specific cycle lie outside the vendor's generic analysis, Duke will reperform the fuel rod thermal analysis to ensure that the results remain bounded by the results of the vendor's generic analysis. In the very unlikely event that the cycle specific thermal analysis results (fuel temperature and pin pressure) are more limiting than the vendor's generic analysis, either the fuel cycle design must be modified or the vendor must resolve the concern within the vendor's ECCS analysis. Responsibility for identification of incompatibility and resolution lies with Duke.

4 i

l 4-5

TABLE 4-1 FUEL MECHANICAL PERFORMANCE ASSESSMENT CRITERIA Analysis Category Linear ECCS Heat Analysis Item Parameter Cladding Cladding Cladding Pin Rate Interface No. Reviewed 1 Collapse Strain 2 Stress Pressure Capability Crit.

1 Pin Power No No No Figure 4-1 Figure 4-1 Figure 4-1 History vs Burnup 2 Radial Assembly Power History vs Burnup Figure 4-2 No No No No No 3 Pellet / Clad No No No No Yes Yes' Gap 4 Clad Yes No Yes No No No Thickness 5 Clad Initial Yes No No No No No Ovality 6 Initial Yes No Yes Yes No Yes Prepressure 7 Densification Yes No No Yes Yes Yes Kinetics NOTES: 1. These criteria are the more significant items reviewed for a reload fuel cycle design, and do not include minor assumptions that are part of the bases.

2. A conservative bounding analysis has been completed which requires no significant assessment for each reload.

4-6

Figure 4 -1

! GENERIC PIN POWER ENVELOPE

! 1.7 a:

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Figure 4-2 GENERIC ASSEMBLY POWER ENVELOPE 15 -

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, :1 t j l 1, i i , , , , i j 0 5000 10000 15000 20000 25000 30000 35000 BURNUP EFPH i

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4-8

Figure 4-3 PIN PRESSURE HISTORY 2400 -

System Pressure 2200 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

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1400 i i , , , , i i C 5000 10000 15000 9nnnn 25000 30000 35000 40000 45000 50000 '

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4-9 1

T Figure 4-4 PLOT OF LHRTM VS. BURNUP 25 -

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i t

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5. MANEUVERING ANALYSIS 5

Section 5 of NFS-1001A discusses the maneuvering analysis used for Oconee reload design. The maneuvering analysis methodology is applied in the same manner when CASMO is used as the lattice code.

f 1

J d

=

j

~' ._ ____. _ . _ .

f 5-1

6. THERMAL hYORAULIC DESIGN Section 6 of NFS-1001A discusses the thermal-hydraulic analysis methodology used for Oconee reload design. The only changes in the thermal-hydraulic methodology as a result of using the CASMO code are the hot channel factors given in Section 6.8.3 of NFS-1001A. Using CASMO, the ratio of the total nuclear uncertainty (1.065) to the radial nuclear uncertainty (1.048) is 1.016. Thus, in determining the generic DNBR curves, F q" changes from 1.065 I to 1.057 when using CASMO.

, Beginning with Oconee 1 cycle 9, Oconee reload cores have contained B&W Mark-BZ fuel. The thermal-hydraulic methodology given in NFS-1001A is used to analyze Mark-BZ reload designs with the changes noted below.

The previously approved BWC critical heat flux correlation (References 7 and

8) is used for Mark-BZ fuel DNBR calculations. The minimum DNBR for Mark-BZ i fuel during steady-state operation, normal operational maneuvers, and anticipated transients is limited to 1.18 using the BWC correlation.

l The rod bow penalty discussed in Section 6.10 of NFS-1001A is no longer applied. A B&W topical report (Reference 9) discusses the mechanisms and resultant effects of fuel rod bow and has been approved by the NRC. This topical report concludes that the ONBR penalty due to rod bow is insignificant and unnecessary because the power production capability of the fuel decreases with irradiation. The empirical correlation developed to predict rod bow also conservatively predicts the rod bow behavior of Mark-BZ fuel (Reference 10).

6-1

A different flux depression hot channel factor (section 6.8.3 of NFS-1001A) is used to develop the generic DN8R curves for Mark-BZ fuel. A 1.007 factor is applied to the calculated axial power since flux depressions at the Zircaloy spacer grids are ignored. Thus, in determining the generic DNBR curves for Mark-BZ fuel, F "q changes from 1.065 to 1.037 when using CASMO.

i i

6-2 4

- -. . - - - - , - - , , .-----,,,.-.---.nn. - , - - . .,n-.----. - ---.--.-. ,- - -,,-._n --,-w.,-,, -, n-- ,.---,,,w.----n,

7. TECHNICAL SPECIFICATIONS REVIEW AND DEVELOPMENT Section 7 of NFS-1001A discusses the process involving the review and development of Technical Specifications for Oconee reload design. Except for minor revisions discussed below, the methodology described in NFS-1001A remains unchanged. As described in Section 6, the BWC critical heat flux (CHF) correlaticn is used for Mark-BZ fuel DN8R calculations. The minimum DN8R for Mark-BZ fuel during steady-state operation, normal operational maneuvers, and anticipated transients is limited to 1.18 using the BWC CHF correlation. As reported and approved in Reference 2, the uncertainty allowances used in Reactor Protection System (RPS) setpoint calculations have been revised to ensure RPS actuation prior to exceeding any safety limits.

The use of the CASMO code and related revisions to nuclear uncertainty factors result in the following changes to specific subsections in NFS-1001A:

A. Section 7.2.2.1 The power power imbalance limits based on the center fuel melt criterion are determined as reported in NFS-1001A. When CASMO is the lattice code upon which the core models are based, a new set of uncertainty factors is applied.

The calculated maximum total peaking factors from the maneuvering analysis are increased by the uncertainty to obtain the 95/95 expected total peaking factor corresponding to each core condition. The individual components of the uncertainty factor are as follows:

7-1

1. Nuclear uncertainty factor = 1.065

a. Bias = 1.63%

b. 95/95 Total Peaking Uncertainty = 4.29%

c. Bounding Radial-Local Uncertainty = 2.2%

2. Engineering hot channel factor = 1.014

3. Densification power spike factor: varies with axial location of the peak in the core. As an upper bound, an 8% uncertainty will be applied for all axial locations.

4. Radial-local factor: varies with location of the assembly in the core (typical value is 1.10). (This is not an uncertainty factor but a direct multiplier on the maximum total peak. This is calculated by the \-core PDQ07 2-D model.)

The nuclear uncertainty factor accounts for the uncertainty in the calculated peak due to the limitations of the analytical models and the radial-local factor is applied to account for the fact that the calculations are performed using an assembly-by-assembly model rather than by using a pin-by pin model.

The engineering hot channel factor accounts for the manufacturing tolerances of critical fuel rod design parameters (pellet enrichment, pellet density, pellet diameter,etc.). The densification power spike factor accounts for the local flux enhancement resulting from gaps in the fuel column induced by fuel l

densification. Althoagh fuel rod bowing is considered to have the potential for enhancing the power peaks, no explicit allowance is provided for the rod 7-2

I bow power spike factor on the basis that the other conservatism factors (nuclear uncertainty factor and engineering hot channel factor) are adequate to offset the effect of the rod bow power spike factor without an additional allowance.

All these factors are statistically combined as follows:

SCUFCFM = 1 + .0163 + 4 0429' + .022' + .014' + .08' = 1.111 where SCUF CFM is the centerline fuel melt statistically combined uncertainty factor.

Limits are established on acceptable valuas of power peaking conditions for each power level and corresponding reactor trip setpoints are established as in Reference 1.

B. Section 7.2.2.2 Sect: 7.2.2.2 of NFS-1001A discusses the DNBR margin calculation. The basic methodology is unchanged by the use of CASMO but the radial uncertainty factor is 1.048 for calculations based on the CASMO lattice code.

C. Section 7.4.1 Power distribution limits based upon ECCS analysis values of maximum allowable linear heat rate are datermined as reported in NFS-1001A. Typical values of the allowable linear heat rates established by the currently applicable ECCS analysis for Oconee cless reacters are gi . m in Table 7-1.

7-3

The effect of non-equilibrium xenon conditions on peaking factors is quantified by the analysis of the power peaking factors occurrina during various power maneuvers. Power redistribution caused by the xenon transient in the power maneuver leads to peaking and offsets being explicitly accounted for in the setting of LOCA limits.

The maneuvering analysis establishes the relationship of operating peaking fac-tors at various axial locations with the core imbalance and control rod posi-tions. The calculated maximum peaks at each axial plane are increased to obtain the 95/95 peaking. The components of the uncertainty factor are as follows:

4

1. Nuclear uncertainty factor = 1.065

a. Bias = 1.63%

b. 95/95 Total Peaking Uncertainty = 4.29%

c. Bounding Local Radial Uncertainty = 2.2%

2. Power Level Uncertainty = 2%

3. Engineering hot channel factor = 1.4%

These factors are statistically combined as follows:

SCUFLOCA , 1 + .0163 + 4.0429z + .022z + .0zz + .014z = 1.070 LOCA where SCUF is the L0i'A statistically ccmbined uncertainty factor. In addition, the radial-local factor is applied to account for local pin peaking.

The radial-local factor varies with location of the assembly in the core (typical value is 1.10).

7-4

Table 7-1 Typical LOCA Kw/ft Limits for Oconee Elevation, 0-1000 1000-2600 After 2600 ft mwd /mtU mwd /mtU mwd /mtU 2 13.5 15.0 15.5 ,

4 16.1 16.6 16.6 6 16.5 18.0 18.0 8 17.0 17.0 17.0

. 10 16.0 16.0 16.0 l

l 1

I l

I r

l 7-5

8. ACCIDENT ANALYSIS REVIEW Section 8 of NFS-1001A discusses the safety analysis methods used in Oconee reload design. The average fuel temperature in Section 8.3.13 is revised to 2950*F as a result of the changes discussed in Section 4. The methodology described in NFS-1001A does not change due to use of the CASMO code.

t l

l f

8-1

9. DEVELOPMENT OF CORE PHYSICS PARAMETERS Section 9 of NFS-1001A discusses the method used to provide core physics data required for startup tests predictions and the core physics report. These methods are applied for the CASMO reload design methodology.

1 I

9-1

10. REFERENCES

1. Duke Power Company Oconee Nuclear Station Reload Design Methodology NFS-1001A April, 1984.

2. Oconee Unit 2, Cycle 7 Reload Report OPC-RO-2002 September 1983.

3. Oconee Unit 2, Cycle 8 Reload Report OPC-RD-2004 December, 1984.

4. Oconee Unit 3, Cycle 7 Reload Report OPC-RO-2001, Revision 1, July 1982.

5. Oconee Unit 3, Cycle 8 Reload Report OPC-RD-2003 February 1984.

6. TACO 2 - Fuel Pin Performance Analysis, B&W-10141PA, Rev. 1, Babcock &

Wilcox, Lynchburg, Virginia, June, 1983.

7. R. H. Wilson, D. A. Farnsworth, and R. H. Stoudt, Correlation of 15 x 15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation, BAW-10143P, Part 2, Babcock & Wilcox, Lynchburg, Virginia, August 1981.

8. J. H. Taylor (B&W) to C. O. Thomas (NRC), Letter, " Acceptance for Referencing of Licensing Topical Report BAW-10143P, Parts 1 and 2", August 2, 1984.

9. J. C. Moxley, et al., Fuel Rod Bowing in Babcock & Wilcox Fuel Designs, BAW-10147P-A, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, May 1983.

10-1

W

10. .J. F. Stolz (NRC) to R. J. Rodriguez (SMUD), letter " Rancho Seco Nuclear Generating Station - Evaluation of Mark-BZ Fuel Assembly Design", November 16, 1984.

11. H. B. Tucker to J. F. Stolz, Oconee Nuclear Station Docket Nos. 50-269,

-270, -287, March 8, 1985.

12. Oconee Unit 3, Cycle 9 Reload Report OPC-RD-2005 June 1985 1 4

10-2

Appendix A r

CODE

SUMMARY

[

i

CASMO CASMO is a multigroup, two-dimensional transport theory code for. burnup calculations on BWR or PWR assemblies. The -2E version used at Duke Power Company is capable of editing the 2 group cross sections by region for P0Q, and the assembly averaged physics data required by EPRI-NCDE-P (Ka, M , Xe/I 2

yields, etc).

CHART CHART creates P0Q tablesets from the CASMO file of cross sections, nuclides, concentrations, and other physics data. Tablesets for the fuel, lumped burnable poison, instrument tube, interassembly gap, moderator filled control rod guide tubes, and control rod regions are prepared for the quarter core PDQ' model.

CASTHF CASTHF, in conjunction with CHART, provides the additional tablesets required to use the thermal feedback option in the Duke Power version of PDQ07. Cross sections as a function of moderator and fuel temperature from CASMO are formatted by CASTHF to supplement the base cross sections from CHART.

EPRI-CELL EPRI-CELL computes the space, energy, and aurnup dependence of the neutron spectrum within cylindrical cells of light water reactor fuel rods. Its A-1

primary output consists of broad group, microscopic, exposure dependent cross sections for subsequent use in multidimensional diffusion theory depletion l analysis. EPRI-CELL uses the methods of three industry accepted subcodes:

GAM-1, THERMOS, and CINOER.

EPRI-N00E-P EPRI-NCDE-P is a 3-0 nodal code similar in theory to FLARE. The Duke Power version of EPRI-N00E-P computes the core effective multiplication factor, the 3-0 power distribution, core coolant flow and teinperature distribution, and fuel exposure distribution. The program includes the effects of partially inserted full length control rods, part length rods, and up to 25 different fuel compositions of varying enrichment and burnable absorber shim loadings.

EPRI-N0DE-P can model a core with up to 48 axial nodes for each fuel assembly and 30 x 30 noces in the XY plane.

The program iterates to account for.the feedback mechanism of moderator and fuel temperature, xenon, and soluble boron on the power distribution and multiplication factor. The program computes the time dependence of iodine and xenon following changes in e.g. power level, power distribution, or control rod movement. Fuel shuffling between cycles and insertion u. new fuel compositions is allowed. Cases may be started for a xenon transient, cycle depletion, or a series of physics branch cases.

A-2

Output includes the 3-0 distributions of power, moderator temperature, iodine and xenon concentrations, burnup, etc. The power offset and imbalance, critical baron concentration, as well as core average moderator and fuel temperature, iodine and xenon concentrations, and burnup are summarized.

EPRI-P0007 MODIFICATIONS PDQ07 is the industry accepted multigroup one, two, or three dimensional diffusion depletion code. The CASMO methodology uses P0Q07/ Version II with modifications to allow options for 20 thermal-hydraulic feedback, improved removal treatment, peak power editing, and re-editing.

NORGE-P NORGE-P formats the lattice physics data calculated by CASMO for input as B-constants to EPRI-N00E-P. Polynomial coef ficients are determined for parameters which vary with temperature and burnup, such as (", M2, xenon reactivity, control rod reactivity, K/v, fI , yy , yxe, etc. NORGE-P automatically plots the input (from CASMO, output) data and the fitted data as well as comparing input to fitted data differences to ensure a good fit.

A-3

4 i

OUKE POWER ~ COMPANY-OCONEE NUCLEAR STATION

!l

.i l PHYSICS TEST COMPARISONS

~

FOR THE CASMO METH000 LOGY 1

i

]

i i

i Technical Report OPC-NE-1002 Supplement 1

March 1985 t

i t

i I'

.yw- - - , - -

l

-ABSTRACT l

t Measurement and calculational techniques and comparisons of calculated and measured results for core physics parameters are presented in this supplement.

The measurements are from Oconee Unit 1, Cycles 7 and 8 and the calculations are performed with EPRI-NODE-P using CASMO 1attice physics data. Comparisons of calculated and measured parameters show good agreement and confirm the adequacy of this calculational procedure in predicting core physics parameters.

I G

S1-ii

TABLE OF CONTENTS

1. Introduction 51 1-1

2. Critical Boron Concentrations 51 2-1 2.1 Measurement Technique S1 2-1 2.2- Calculational Technique 51 2 2. 3 ' Comparisons of Calculated and Measured Results S1 2-2 2.4 Summary S1 2-3

3. -Control Rod Worths S1 3-1 3.1 Measurement Techniques S1 3-1 d

3.2 Calculational Techniques 51 3-1 3.3 Comparisons of Calculated and Measured Results 51 3-2 3.4 Summary. 51 3-3

a. Ejected Rod Worths S1.4-1 1

5. . Isothermal Temperature Coefficients S1 5-1 5.1 Measurement Technique S1 5-1 5.2 Calculational Technique S1-5-1 5.3 Comparison of Calculated and Measured Results S1 5-2 5.4 Summary S1 5-2

6. References S1 6-1

S1-iii

LIST OF TABLES Page 2-1 Oconee 1 Cycles 7 and 8 BOC Critical Boron Concentrations at Hot Zero Power 51 2-4 2-2 Oconee 1 Cycles 7 and 8 Hot Full Power Critical Boron Concentrations from EPRI-NODE-P vs. Measured S1 2-5 2-3 Oconee 1 Cycles 7 and 8 Hot Full Power Critical Boron Concentrations from Quarter Core, Fine Mesh PDQ vs. Measured S1 2-7 3-1 Oconee 1 Cycles 7 and 8 Control Rod Worths At Hot Zero Power, B0C In Terms of Reactivity S1 3-4 3-2 Oconee 1 Cycles 7 and 8 Control Rod Worths At HZP, B0C In Terms of Baron S1 3-5 1

5-1 Oconee 1 Cycles 7 and 8 Isothermal Temperature Coefficients at HZP, 80C S1 5-3 SI-iv

LIST OF FIGURES P, age 2-1 Oconee 1, Cycle 7 - Baron Letdown - HFP, ARO, APSR IN, NODE vs. Meas Baron 51 2-8 2-2 Oconee 1, Cycle 8 - Boron Letdown - HFP, ARO, APSR IN, NODE vs. Meas Boron S1 2-9 l

S1-v i

1. INTRODUCTION This supplement presents measurement and calculational techniques and compar-isons of calculated and measured results for some key core physics parameters.

The physics parameters include het zero power (HZP) and hot full power (HFP) critical baron concentrations, HZP control rod worths and ejected rod worths, and HZP isothermal temperature coefficients.

The measured data is from the Oconee Nuclear Station Unit 1, Cycles 7 and 8.

The measurement techniques discussed are those currently used at the station.

The HZP measurements were taken at beginning-of-cycle (80C) during the Zero Power Physics Testing. The HFP boron concentration measurements were taken at various time steps throughout the cycles.

All calculations were performed with EPRI-NODE-P. In contrast to predictions, which are calculated before the measurements are taken, the calculations pre-sented here were performed after the measurements were taken. Therefore, the plant conditions at the time of the measurements could be closely modeled with EPRI-N0DE-P.

The comparisons of calculated and measured results present the means of the differences between the measured and calculated data and the corresponding standard deviations. The mean and standard deviation are defined as follows:

S1 1-1

Ix.

Mean = i = n Standard _ _1

' *i}

Deviation - n-1 where: x 9 = value for the ith observation I

n = number of observations.

1 51 1-2

2. CRITICAL BORON CONCENTRATIONS 2.1 Measurement Technique Critical boron concentrations are measured at HZP and HFP by an acid-base

~ titration of a reactor coolant system sample.

The measurement uncertainty for critical baron concentrations is due to (1) error in the titration method and (2) error due to differences between the sample concentration and the core average concentration. Based on conservative estimates of these errors, the total uncertainty associated with the critical boron concentration measurements is less than 20 ppmb.

2.2 Calculational Technique Critical boron concentrations are calculated at HZP and HFP using EPRI-NODE-P in the boron search mode. The search can get boron within 10.5 PPM of critical; if any correction to critical is needed, a boron worth calculated by EPRI-N0DE-P is used.

The quarter core, fine mesh P0Q model also provides a critical boron concentration. Using the input boron concentration and the resulting eigen-value, for each timestep the exactly critical (A = 1.0) boron is determined using a boron worth calculated by P0Q or EPRI-N0DE-P.

l 51 2-1

2.3 Comparison of Calculated and Measured Results l

2.3.1 Hot Zero Power Comparison l

l The calculated and measured critical baron concentrations at HZP and BOC for Oconee Unit 1, Cycles 7 and 8 are compared in Table 2-1. One value per cycle, at HZP, ARO, APSR centered, are shown since measurements are no longer taken at control rod endpoints as for cycles 1 thru 5 (Reference 1, page 51 2-1).

2.3.2 Hot Full Power Comparison The calculated (from EPRI-NODE-P) and measured critical baron concentrations at HFP for cycles 7 and 8 of Oconee Unit 1 are compared in Table 2-2. The mean of the differences for both cycles is -17 ppmb with a standard deviation of 13 ppab.

The quarter core P0Q critical baron is compared to the measured baron adjusted to all rods out (except group 8), equilibrium xenon, and other factors which may be off-nominal. PDQ to measured HFP critical bcron concentrations are compared in Table 2-3. The mean of the differences for both cycles is -16 ppmb with a standard deviation of 21 ppmb.

The data displayed in Table 2-2 can be visualized better by examining plots of soluble boron concentration as a function of burnup. These boron letdown curves are shown in Figures 2-1 and 2-2.

51 2-2

2.4 - Summary The comparison between EPRI-N00E-P and measured critical boron concentrations at HZP and HFP indicate that the CASMO based EPRI-NODE-P can adequately predict soluble boron concentrations.

i i

6 4

! S1 2-3

Table 2-1 Oconee 1 Cycles 7 and 8 BOC Critical Baron Concentrations at Hot Zero Power Groups 1-7 out, Group 8 at 37.5 % WD Critical Baron Conc. PPM Difference, Cycle Calculated Measured PPM 7 1618 1622 -4 8 1554 1577 -23 Mean -13.5 Standard Deviation 13.4 4

1 I

51 2-4 I

l

Table 2-2 Oconee 1 Cycles 7 and 8 Hct Full Power Critical Boron Concentrations From EPRI-N0DE-P vs. Measured.

Critical Baron, PPM Difference, Cycle EFPD Calculated (1) Measured PPM 7 6.94 1124 1140 16* -

14.84 1115 1130 15*

19.80 1103 1060 -43 36.04 1074 1040 -34 48.73 1048 1020 -28 53./3 1037 1010 -27 66.91 1016 988 -28 74.57 991 978 -13 84.54 968 944 -24 93.47 948 866 -82*

107.84 914 890 -24 119.36 885 863 -22 -

142.79 826 809 -17 160.33 781 770 -11 171.33 752 744 -8 178.33 734 705 -29 191.57 699 685 -14 208.95 651 618 -33 223.95 610 580 -30 238.92 567 542 -25 254.78 523 494 -29 271.02 478 449 -29 285.06 439 411 -28 300.05 397 365 -28 312.88 358 332 -26 327.29 319 297 -22 342.83 274 247 -27 357.85 232 202 -30 372.80 190 159 -31 387.78 147 118 -29

. 399.81 113 89 -24 402.87 156 105 -51" (continued) 51 2-5

f Table 2-2 (Continued)

Critical Boron, PPM Difference, Cycle EFPD Calculated (1) Measured PPM 8 4.05 1053 1176 123*

7.04 1048 1030 -18 9.02 1047 1035 -12 21.02 1030 991 -39 37.92 998 964 -34 52.93 966 949 -17 68.94 930 926 -4 82.96 901 899 -2 99.11 864 868 4 113.83 830 835 5 129.84 791 794 3 145.71 752 760 8 160.74 715 724 9 172.77 684 659 -25*

189.67 639 639 0 203.33 605 606 1 219.27 563 560 -3 234.37 522 523 1 '

249.27 483 480 -3 263.59 444 452 8 278.38 405 394 -11 294.29 368 351 -17 309.30 320 310 -10 324.26 280 268 -12 340.22 235 218 -17 355.20 193 176 -17 370.26 153 137 -16 385.25 113 94 -19 400.26 72 48 -24 Mean -17 Standard Deviation 13 Difference = Measured - Calculated

! (1) EPRI-NODE-P calculated boron at conditions as close to the measured conditions as possible. This is then the results of the core follow, rather than nominal, calculation.

(*) These points are omitted from the mean and standard deviation calculation.

r t

51 2-6 l

Table 2-3 Oconee 1 Cycles 7 and 8 Hot Full Power Critical Boron Concentrations From Quarter Core, Fine Mesh PDQ vs. Measured All rods out, Nominal Critical Boron, PPM Difference, Cycle EFPD Calculated Measured PPM 7 4 1130 1115 -15 12 1108 1086 -22 25 1082 1063 -19 50 1042 1017 -25 100 944 915 -29 150 826 793 -33 200 697 657 -40 250 558 515 -43 300 415 370 -45 350 268 227 -41 403 111 75 -36 8 4 1022 1030 8 12 997 1005 8 25 973 990 17 50 937 958 21 100 848 866 18 150 739 749 10 200 620 618 -2 250 494 481 -13 300 361 340 -21 350 225 201 -24 400 82 61 -21 Mean -16 Standard Deviation 21 1

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3. CONTROL R00 WORTHS 3.1 Measurement Techniques Individual control rod group worths are measured by the boron swap technique.

This technique involves a continuous decrease in baron concentration together with an insertion of the control rods in small, discrete steps. The change in reactivity due to each insertion is determined from reactimeter readings before and after the insertion. The worth of each rod group is the sum of all the reactivity changes for that group.

The worth of the total regulating banks 5-7 can be measured in two ways. The first is to add up the worths of the individual banks as determined from the reactimeter readings. The worth in this case is in terms of reactivity. This measurement technique is the approved test procedure method. The second way is to measure rod worth in terms of change in soluble boron concentration. This worth is defined e.s the difference between the equilibrium critical boron concentration when all the regulating rods are out and the concentration when all the regulating rods are in.

3.2 Calculational Techniques Individual and total regulating rod group worths in terms of reactivity are calculated by making two EPRI-NODE-P runs. The first is a boron search run with the rod group (s) out. The boron concentration found in this run is then used in a fixed baron run with the rod group (s) in. The difference in reac-tivity between these two runs with constant boron concentration is the rod group (s) worth.

51 3-1

l To calculate the total worth of the regulating rod groups in terms of soluble boron concentration, a baron search to critical using EPRI-N0DE-P is performed l

l for both the rods-cut case and the rods-in case. If necessary, the resulting j

boron concentrations are then corrected to exact criticality. Finally, the group worths are determined as the difference between these critical boron concentrations.

3.3 Comparison of Calculated and Measureo Results -

3.3.1 Comparison in Terms of Reactivity A comparison of calculated and measured control rod worths in terms of reac-tivity is shown in Table 3-1. This table compares the worths of the individual banks 5, 6 and 7 and the total regulating banks 5-7 at HZP and B0C for Oconee Unit 1, Cycles 7 and 8. The differences between measured and calculated worths for all the banks are small. For the total banks 5-7, the mean of the percent difference was 2.5% with a standard deviation of 1.0%.

3.3.2 Comparison in Terms of Soluble Boron Concentration The comparison of calculated and measured regulating bank worths in terms of soluble boron concentration for Oconee Unit 1, Cycles 7 and 8 at HZP and BOC is given in Table 3-2. The agreement between calculated and measured worths is very good. All the differences are less than the measurement uncertainty associated with the boron measurement. The mean of the differences between measured and calculated values for the total banks 5-7 was found to be -4 ppmb with a standard deviation of 4 ppmb. The mean of the % differences was -1.08 with a standard deviation of 1.16.

51 3-2

(

3.4 Summary The comparisons between the calculated and measured control rod worths at HZP indicate that the CASMO based EPRI-NODE-P can adequately predict control rod worths. This has been verified by comparing calculated worths to two independent measurement techniques for Oconee 1, Cycles 7 and 8.

S1 3-3

Table 3-1 Oconee 1 Cycles 7 and 8 Control Rod Worths at Hot Zero Power, 80C In Terms of Reactivity Rod Worth, % ap Difference, ,

Cycle Bank Calculated Measured  % ap  % Difference l 7 7 1.23 1.34 0.11 8.2 6 0.81 0.93 0.12 12.9 5 1.28 1.18 -0.10 -8.5 5-7 3.33 3.44 0.11 3.2 ,

8 7 1.29 1.35 0.06 4.4 6 0.84 0.86 0.02 2.3 5 1.08 1.07 -0.01 -0.9 5-7 3.22 3.28 0.06 1.8 Mean 7 NA NA 0.085 6.3 6 NA NA 0.070 7.6 5 NA NA -0.055 -4.7 5-7 NA NA 0.085 2.5 Standard 7 NA NA 0.035 2.7 Deviation 6 NA NA 0.071 7.5 5 NA NA 0.064 5.4 5-7 NA NA 0.035 1. 0 Group 8 at 37.5% WD Difference = Measured - Calculated

% Difference = Difference Measured

• 100 51 3-4

r

(

I l

i Table 3-2 i Oconee 1 Cycles 7 and 8 Control Rod Worths at HZP, BOC in Terms of Baron l i

Rod Worth, PPM Difference, Cycle Banks Calculated Measured PPM  % Difference s

7 5-7 380 379 -1 -0.26 8 5-7 376 369 -7 -1.90 Mean -4 -1.08 Standard Deviation 4 1.16 f

f 51 3-5 i

4. Ejected Rod Worths

- This section deleted -

51 4-1 REV 1

l

[

4.1.3 Rod Drop The rod drop method starts with-the ejected rod fully withdrawn. The rod is then tripped into the core and the reactivity is charted by the reactimeter.

Rod position and boron concentration are held constant. The extrapolation to zero inverse time of a plot of reactivity vs. inverse time yields the ejected rod worth. The uncertainty associated with this method is mt.ch greater than that associated with either of the other two methods.

4.2 Calculational Techniques Ejected rod worths are calculated using EPRI-N00E-P to simulate either boron swap, rod swap, or rod drop.

4.2.1 Boron Swap For boron swap, a boron search run is first performed to determine the critical boron concentration at the rod group position. The baron concentration as calculated in the EPRI-NODE-P run should be corrected for exact , criticality.

Using this corrected baron concentration and a constant rod group position, the reactivity is determined with the worst case rod first in and then out. The ejected rod worth is the difference in reactivity between the worst case rod in and out.

4.2.2 Rod Swap For rod swap, the reactivity is determined with the worst case rod first in and then out, keeping the boron concentration constant and the controlling rod S1 4-2

group position constant at the least withdrawn position (least withdrawn corresponds to ejected rod out). The ejected rod worth is the difference in reactivity between the worst case rod in and out.

4.2.3 Rod'Orop To calculate ejected rod worths by rod drop, EPRI-NODE-P cases with the worst case rod first out and then in are executed, with boron concentration and rod group position held constant. The ejected rod worth is the difference in reactivity between these two cases.

4.3 Comparison of Calculated and Measured Results A comparison of calculated and measured ejected rod worth for Oconee Unit 1, Cycle 7 is given in Table 4-1. The agreement is good. The difference between the measured and calculated values is 0.06 % ak/k.

4.4 Summary The comparison between measured and calculated ejected rod worths indicate that the CASMO based EPRI-NODE-P can adequately predict ejected rod worths.

51 4-3

Table 4-1 Oconee 1 Cycle 7 Ejected Rod Worth Measurement Worth, % AP Difference Location Technique Calculated Measured  % ao N-12 Baron Swap 0.35 0.41 0.06 1

l j

i 4

51 4-4

m L.

l 5. ISOTHERMAL TEMPERATURE COEFFIC:ENTS i

f .

l The isothermal temperature coefficient is defined as the change in reactivity 1

per unit change in moderator temper :ture at hot zero power, i.e. ,

"T

• 5.1 Measurement Technique The isothermal temperature coefficient is measured by executing average moderator temperature changes from a plateau of initial equilibrium critical conditions. After the change, steady state conditions are established and pertinent data are recorded by the reactimeter or reactivity program at the resulting plateau. The isothermal temperature coefficient is determined as the

- change in reactivity between plateaus divided by the change in temperature.

The uncertainty associated with the isothermal temperature coefficient measure-ment is dependent on the value of the coefficient. For acceptable coefficient

~

values, the measurement uncertainty is less than 10.06 x 10 4 Ak/k/ F.

5.2 Calculational Technique 4

The isothermal temperature coefficient at HZP is calculated using EPRI-N00E-P.

Two cases with the same baron concentration and rod positions but different I

~

moderator temperatures are run. The isothermal temperature coefficient is the difference in reactivity between the two cases divided by the difference in the moderator temperatures.

i l 51 5-1

5.3 Comparison of Calculated and Measured Results A comparison of calculated and measured isothermal temperature coefficients at HZP and BOC for Oconee Unit 1, Cycles 7 and 8 is presented in Table 5-1. The agreement between these calculated and measured coefficients is very good; all

~

values are well within 0.3 x 10 4 ap/*F acceptance criteria. 1 5.4 Summary The comparison between calculated and measured isothermal temperature coef-ficients indicates that the CASMO based EPRI-N0DE-P is an adequate predictor of '

isothermal temperature coefficients.

l l

S1 5-2 1

Table 5-1 Oconee 1 Cycles 7 and 8 Isothermal Temperature Coefficients at HZP, BOC Boron Conc, Temp. Coeff, 10 4 ap/'F Difference Cycle PPM Calculated Measured 10 d ap/*F 7 1610 +0.229 +0.273 +0.044 8 1635 +0.08 +0.11 +0.03 1

S1 5-3 i

6. REFERENCES

1. Oconee Nuclear Station Reload Design Methodology, NIS-1001A, Duke Power Company, April 1984.

S1 6-1

7 DUKE POWER COMPANY OCONEE NUCLEAR STATION NUCLEAR RELIABILITY FACTORS FOR EPRI-NODE-P BASED ON CASMO LATTICE PHYSICS DATA Technical Report OPC-NE-1002 Supplement 2 March 1985

7. -.

4 ABSTRACT Supplement 2 describes Duke Power Company's benchmarking of a CASMO based i

EPRI-NODE-P. Included in this supplement are measured Assembly Powers, Radial-Local Comparisons, EPRI-NODE-P Calculations, Statistical Analyses, and Fitting Procedures.

1.

I i.

i I

a 4

I 52-11 i

TABL 0F CONTENTS 1

1. Introduction and Summary 52 1-1 l

1.1 Introduction 52 1-1

1. 2 Structure of Supplement E S2 1-1 1.3 Summary 52 1-2

2. Measurement Data S2 2-1 2.1 Measured Assembly Power Data 52 2-1 2.2 Measurement System Description S2 2-1

2. 3 Measured Powers: Cycle 7 and Cycle 8 S2 2-2

3. Radial-Local Analysis 52 3-1 3.1 Radial-Local Reliability Factor Analysis 52 3-1 3.2 Comparisons of PDQ to CASMO S2 3-1 3.3 Comparisons of CASMO to Criticals 52 3-4 3.4 Conclusions 52 3-4

4. EPRI-NODE-P Power Distribution Comparisons 52 4-1 4.1 EPRI-NODE-P Model S2 4-1 4.2 Oconee Fuel Cycle Simulations 52 4-2 4.3 PDQ07 MODEL S2 4-3 4.4 Conclusions 52 4-3

5. Statistical Analysis 52 5-1 5.1 Observed Nuclear Reliability Factor Derivation 52 5-1 5.2 Normality Test Results 52 5-2 5.3 Observed Nuclear Reliability Factors For EPRI-NODE-P S2 5-3 5.4 Quantitative Comparisons of EPRI-N0DE-P to Measurements 52 5-4

5. 5 ONRF Confirmation S2 5-5 5.6 Conclusions 52 5-5

6. References 52 6-1 S2-iii

LIST OF TABLES Pag 2-1 Oconee 1 Cycle 7 Failed Detector Strings 52-2-4 3-1 CASMO and PDQ k Assembly Depletion Peak Pin Powers and Differences 52 3-5 3-2 CASMO vs. Experiment - Kritz Critir.als Fission Rate Deviations 52 3-6 4-1 Oconee Unit 1 Cycle 7 State Points 52 4-4 4-2 Oconee Unit 1 Cycle 8 State Points 52 4-5 5-1 Difference Distribution Normality Tests ( (C, M) > 1.0) 52 5-6 2 EPRI-N0DE-P ONRF Calculation 52 5-7 ,

5-3 Difference Means and Standard Deviations for Peaks S2 5-8 5-4 Difference Means and Standard Deviations for Radials 52 5-9 S2-iv

i LIST OF FIGURES Page 2-1 Oconee Instrument String Locations 52 2-5 2-2 Detector String Number Assignment for Eighth-Core Averaging S2 2-6 3-1 Assembly Power Distribution - 0 MWD /MTU S2 3-7 3-2 Assembly Power Distribution - 5000 MWD /MTU S2 3-8 3-3 Assembly Power Distribution - 10,000 MWO/MTU S2 3-9 3-4 Assembly Power Distribution - 20,000 MWD /MTU S2 3-10 3-5 Assembly Power Distribution - 40,000 MWD /MTU S2 3-11 4-1 NODE-P and PDQ Setup Methodology 52 4-6 4-2 BOC Radial Power Distribution - Cycle 7 S2 4-7 4-3 MOC Radial Power Distribution - Cycle 7 52 4-8 4-4 EOC Radial Power Distribution - Cycle 7 S2 4-9 4-5 BOC Total Power Distribution - Cycle 7 S2 4-10 4-6 MOC Total Power Distribution - Cycle 7 S2 4-11 4-7 E0C Total Power Distribution - Cycle 7 52 4-12 4-8 BOC Axial Power in Hot Assembly - Cycle 7 52 4-13 4-9 MOC Axial Power in Hot Assembly - Cycle 7 52 4-14 4-10 EOC Axial Power in Hot Assembly - Cycle 7 52 4-15 4-11 BOC Axial Power in APSR Assembly - Cycle 7 S2 4-16 4-12 MOC Axial Power in APSR Assembly - Cycle 7 52 4-17 4-13 EOC Axial Power in APSR Assembly - Cycle 7 52 4-18 4-14 BOC Radial Power Distribution - Cycle 8 52 4-1?

4-15 MOC Radial Power Distribution - Cyc h 8 S2 4-20 4-16 EOC Radial Power Distribution - Cycle 8 52 4-21 4-17 B0C Total Power Distribution - Cycle 8 52 4-22 52-v

r LIST OF FIGURES (CONTINUED)

PaSe 4-18 MOC Total Power Distribution - Cycle 8 S2 4-23 4-19 EOC Total Power Distribution - Cycle 8 52 4-24 4-20 BOC Axial Power in Hot Assembly - Cycle 8 S2 4-25 4-21 MOC Axial Power in Hot Assembly - Cycle 8 S2 4-26 4-22 EOC Axial Power is Hot Assembly - Cycle 8 52 4-27 4-23 BOC Axial Power in Gad Assembly - Cycle 8 S2 4-28 4-24 MOC Axial Power in Gad Assembly - Cycle 8 S2 4-29 4-25 EOC Axial Power in Gad Assembly - Cycle 8 52 4-30 4-26 BOC PDQ Radial Power Distribution - Cycle 7 52 4-31 4-27 MOC PDQ Radial Power Distribution - Cycle 7 S2 4-32 4-28 EOC PDQ Radial Power Distribution - Cycle 7 52 4-33 4-29 BOC PDQ Radial Power Distribution - Cycle 8 52 4-34 4-30 MOC PDQ Radial Power Distribution - Cycle 8 52 4-35 4-31 EOC PDQ Radial Power Distribution - Cycle 8 52 4-36 5-1 Oconee 1 Cycle 7 Radial Frequency Distribution 52 5-10 5-2 Oconee 1 Cycle 8 Radial Frequency Distribution 52 5-11 5-3 Combined Radial Frequency Distribution 52 5-12 5-4 Oconee 1 Cycle 7 Total Frequency Distributuion 52 5-13 5-5 Oconee 1 Cycle 8 Total Frequency Distributuion 52 5-14 5-6 Combined Total Frequency Distributuion 52 5-15 52-vi

INTRODUCTION AND

SUMMARY

1.1 Introduction The current nuclear code used by Duke Power Company for,three dimensional assembly power calculations is EPRI-NODE-P. In NFS-1001A, this code has been benchmarked (using the EPRI-ARMP methodology to determine the lattice physics data used by NODE) against Oconee Unit 1, Cycles 1 thru 5. The reliability factors for EPRI-NODE-P using lattice physics data generated by CASMO are determined here.

This work encompassed: obtaining measured power distributions for Oconee Unit 1, Cycles 7 and 8, simulations of these 2 cycles using EPRI-N0DE-P, and determination of reliability factors for EPRI-NODE-P (using a CASMO data base) with the statistical methods developed in Section 5 of Appendix A of reference 1.

1.2 Structure of Supplement 2 This supplement considers the parameters required to obtain statistically based reliability factors for EPRI-NODE-P. Section 2 will describe the assembly measured power data base. Section 3 compares calculated and measured radial-local factors. Section 4 describes the EPRI-NODE-P Oconee simulations and presents comparisons of assembly radial and peak powers. Section 5 quantitatively compares calculated and measured powers from Section 4.

52 1-1 l

l

1. 3 Summary Radial-local factors (pin powers within an assembly) predicted by PDQ07 were examined by comparing to an identical CASMO calculation. The PDQ was shown to underpredict the CASMO pin power by a maximum of 1.4%. After statistically combining this PDQ to CASMO error with the CASMO to experiment error of 1.7%,

the radial-local uncertainty for P0Q pin powers is 2.2%.

Calculated and measured powers were statistically compared to derive 95/95 Observed Nuclear Reliability Factors (ONRF) for the CASMO based EPRI-NODE-P.

Using Oconee Unit 1, Cycles 7 and 8, the ONRF's for the assembly radial and assembly peak nodal powers were found to be 1.038 and 1.059 respectively.

Combining the EPRI-N0DE-P ONRFs with the PDQ pin power uncertainty gives the statistically combined uncertainty factors of 1.048 radial and 1.065 total.

52 1-2

2. MEASUREMENT DATA 2.1 Measured Assembly Power Data The measured power data base used in this , supplement comprises assembly power All assembly power data are dats from Oconee Unit 1 for cycles 7 and 8.

directly traceable to raw signals received from the incore detector system.

2.2 Measurement System Description Tha incore detectors at Oconee consist of pure Rhodium emitters which respond ,

With each neutron absorption, a beta particle to the incident neutron flux.

(S) is released according to the reaction:

Rh + 0_ $ + energy (2-1)

Rh+fn+

Th2 current measured from the emitter to ground is proportional to the net minutes),the emitter loss. After the emitter current has stabilized (~4 current is then proportional to the local neutron flux (in the neighborhood of This emitter is called a Self P,owered the eight pins surrounding the emitter).

N;utron Detector (SPND).

The SPN0's are physically located inside the Fuel A,ssembly (FA) Instrument Tube (IT).

The IT is situated in the center of the FA.

The reactor's on-line SPN0 signal magnitude is of the order of nanoamps.

computer (OLC) performs a signal to power conversion at approximately ten minute intervals and logs signal data, core power, power distribution, and assorted other data pertinent to core operation.

52 2-1

SPN0's are distributed in fixed positions to provide an adequate three dimensional assembly power measurement. In each instrumented FA, seven SPND's

! are located equidistantly along a " string." Each string also has a

! thermocouple and an insulated leadwire which is used to correct for gamma

! induced signals in the seven detector leads. ~

i In Oconee, 52 of 177 FA contain detector strings. The locations of instrumented FA form a spiral as shown in Figure 2-1. Eight strings are located symmetrically in the interior; another 8 symmetrically farther out toward the periphery. The two sets of eight strings are used to supply corewise quadrant tilt information. An eighth core map of 29 FA three dimensional powers can be obtained; and using tilt data, full core maps of 177 radial FA powers and 1239 segment FA powers can be obtained.

The measured powers used in this supplement will be collapsed from 52x7 (full core) to 29x7 (eighth core) at each reactor state point. As shown on Figure 2-2, 11 of the 29 eighth core locations have symmetrically located detector strings. Relative powers (radial and seven-level) for each symmetric pair or symmetric octet were averaged to obtain the best estimate of the "true" measured power.

Power measurements were taken at approximately equilibrium xenon conditions.

Reactor power was also as close to 100% full power as practicable.

2.3 Measured Powers: Cycle 7 And Cycle 8 All measured power distributions are from the Oconee Unit 1 Performance Data Output (PDO) compiled by the Operator Aid Computer (OAC). Intervals of 52 2-2

. approximately 2 weeks were chosen for the core follow statepoints. These are l

shown in Tables 4-1 and 4-2.

l No explicit estimate is made here of measurement system accuracy during cycles 7 and 8 since this component is conservatively treated in section 5.

Substitute signals for failed detectors are derived through a spline fitting procedure, provided that operating SPND's are adjacent on either side to the failed SPND. If two or more adjacent SPND's fail on a string, substitute signals are derived from either symmetric or adjacent locations. The same procedure is used for entire string substitution. In cycle 7, 5 assemblies had enough strings being substituted for that they were removed from the reliability calculation. These are shown in Table 2-1. No assemblies were removed from cycle 8 measured data.

4 S2 2-3

Table 2-1 -

Oconee 1 Cycle 7 Failed Detector Strings Assembly String Assembly Number in Number location Eighth Core 2 H-9 2 10 H-5 4 14 N-8 5 6 F-7 10 8 G-6 10 i

i i

l.

52 2-4 L

Figure 2-1

'. c o ne e Instrument String Locations X

i A l B 31 30 C

32+ 29 2S+ 52 D 33 ' 27 51 34 7* 5* 26 7 35+ 6 4 24 23+

g 36 9* 8 3 25* 22 g 37 10 1 2 21 Y 11" 20 I

195-K ,

J5 39- 14 is 50+

L l 40 1.: " low 17 49

• 1 14 15 N

l 42 43+ 47+ ad

., I l 44 i

l 45 4o R

2 l

[

l 2 3 4 5 6 3 9 10 11 . 13 14 13

• - Inner Eignt Symetric Detector 5trings

+ - Outer Eight Symmetric Setector Strings l

i S2 2-5

l Figure 2-2 Oconee Fuel Assembly Map Detector String Number Assignment For Eighth-Co're Averaging l

8 9 10 11 12 13 14 15 H 1 2 4 10 14 21 30 37 0 15 22 31 3 1 13 16 45 8 20 29 36 19 25 17 24 23 28 32 L 38 12 35 39 43 46 18 27 47 50 33 40 M 26 49 34 42

, N 41 48 51 0 52 S2 2-6

3. RADIAL-LOCAL ANALYSIS 3.1 Radial-Local Reliability Factor Analysis In this document, the radial-local is defined as the ratio of the maximum pin power to the assembly radial power (assembly average X-Y' power). Commercial power reactors, such as Oconee, are not instrumented to measure radial-local factors. In order to predict the expected pin peaking for a core design, a 20, quarter core P0Q model, which calculates pin by pin powers throughout the cycle, is used. The variation of P0Q predicted pin power from the true pin power is not known for Oconee cores but can be inferred from comparisons to other experiments where pin powers are measured.

In the ' case of a CASH 0 based P0Q, the P0Q to measured deviation will be.._

^

examined in two parts. First, the PDQ to CASMO radial-local factors for a quarter assembly geometry, typical of the quarter core geometry, will be compared and the error determined. Next, the CASMO radial-local factors are compared to the Kritz hot criticals (Reference 2) experimental results.

Statistically combining these will give an effective PDQ to measured reliability factor.

- 3. 2 Comparisons of PDQ to CASMO The procedure to compare the PDQ and CASMO power distributions is as follows:

1. Following the methodology described in Sections 3.1.1.1 and 3.1.1.2 of this report, a quarter assembly analysis of a typical B&W 15x15 assembly of the Oconee design was performed by CASMO and P0Q. This assembly was S2 3-1

235 loaded with 3.28 w/o U fuel at 95% TD with 16 1.0 w/o B 4C burnable poison rods in the control rod guide tubes. This is typical ~f o the feed batch design at Oconee.

2. The assembly was depleted under the same conditions as in the quarter core P0Q (with the exception that no moderator or fuel temperature feedback is used) and followed a depletion history typically seen.

a. Deplete at HFP conditions to 20,000 MWD /MTU with the burnable poison in.

b. Remove burnable poison at 20,000 MWD /MTU and deplete with moderator filled guide tubes to 40,000 MWD /MTU.

3. The pin power distribution throughout the depletion was analyzed. For all pins where CASMO predicts a power greater than 1.00, determine the difference between the PDQ and CASM0 pin powers. This selection is made for the following reasons:

a. The peak pin power only is used in the margin to centerline fuel melt and LOCA in the maneuvering analysis. Thus, those pins of low power are of no concern.

b. Just the peak pin cannot be evaluated since the hot channel (the power of four adjacent pins) is considered in the generic DNB calculation. Therefore, the pins at less than peak power must be examined for po;sibly larger deviations.

52 3-2

f I

c. PDQ may predict the hot pin to be in a different location than does CASMO. Therefore, the PDQ power in the same location as CASMO must be considered in the analysis.

The results at several selected timesteps are shown in Figures 3-1 thru 3-5. Table 3-1 lists the peak pin power at all CASMO timesteps for CASM0, PDQ, and the difference (PDQ-CASMO). It is seen that at all but the first step PDQ underpredicts the peak pin. There are several reasons for this.

PDQ, a oiffusion theory code, cannot model the 2 group fluxes to the extent required. (CASM0, a transport theor/ code, can and also performs the 2D calculation in 7 groups.) This is most evident when burnable poison depletion sets in and the power (flux) is not as depressed in the fuel surrounding the depleted BP. The BP cross sections in PDQ have been adjusted by CASMO, which calculated all fuel and non-fuel cross sections for this P0Q, to account for the diffusion theory approximation. The diffusion corrected BP cross sections tend to raise the flux in the area of the BP and thus lowering it elsewhere. While this is necessary for the proper depletion of the BP and overall power distribution, the effect is to decrease the power in high power pins.

4. The differences were examined for normality and found not to be normally distributed. Therefore, a conservative upper bounding deviation will be used. The maximum deviation in pin powers between PDQ and CASMO is 1.4%

, at 40,000 MWD /MTU assembly burnup. Since this high burnup is greater than any limiting assembly burnup (in most cases it is new fuel which encounters the maximum power peaking), the 1.4% upper bound will be conservatively applied to all core locations.

52 3-3

! 3.3 Comparisons of CASMO to Criticals I An earlier version of CASMO has been banchmarked against the hot criticals at the Kritz facility. The version of CASMO used at Duke Power has improved l flexibility and has certain edit capabilities not fount in the early version.

The transport theory calculations are the same in each version and hence the benchmarking discussed in Reference 2 applies to the CASMO used at Duke Power.

The results of the two PWR assembly experiments are given in Table 3-2. These experiments were performed with lattice configurations and dimensic1s which

• are similar to standard PWR assemblies. The actual dimensions are proprietary and have not been published.

The results of the two PWR experiments are combined to better represent assembly configurations present in Oconee cores. Current cycles contain assemblies which contain lumped burnable poisons as well as assemblies without burnable poisons. The Kritz experiment with burnabie poisons was performed at less than 100 PPMB and the other experiment was performed at approximately 1000 PPMB. This range of boron concentrations is representative of Oconee fuel cycles. The results are given in Table 3-2 and give a CASMO to experiment deviation of 1.7%.

S2 3-4A REV 1

3.4 Conclusions The bounding PDQ to CASMO pin power difference and the CASMO to experiment error can be statistically combined to give the expected radial-local uncertainty factor for CASMO based reload design. This factor is calculated as:

FR-L = 4 0142 + .017z = 0.022 This 2.2% conservatively accounts for pin power error from the quarter-core PDQ07 model.

e S2 3-4B REV 1

(

Table 3-1 CASMO and PDQ k Assembly Depletion Peak Pin Powers and Differences Assembly Peak Pin Powers Difference Burnup, from P0Q -

NWD/MTV CASMO PDQ CASMO O 1.091 1.091 0.0 100 1.092 1.089 -0.003 500 1.091 1.085 -0.006 1000 1.088 1.080 -0.008 2000 1.078 1.070 -0.008 3000 1.069 1.061 -0.008 5000 1.054 1.046 -0.008 7500 1.038 1.031 -0.007 10000 1.029 1.027 -0.002 12500 1.026 1.024 -0.002 15000 1.026 1.022 -0.004 20000 (a) 1.027 1.021 -0.006 20000 (b) 1.074 1.070 -0.004 25000' 1.063 1.055 -0.008 30000 1.053 1.043 -0.010 35000 1.045 1.033 -0.012 40000 1.038 ,1.024 -0.014 (a) BP Rods in to this point.

(b) BP Rods pulled here and the depletion continued.

R-52 3-5

Table 3-2 CASH 0 vs. Experiment - Kritz Criticals Fission Rate Deviations Deviation = 100 * (CASMO-EXPERIMENT)/ EXPERIMENT 15x15 PWR M0 2 14x14 PWR M0 2 with with Absorbers and Water Holes Water Holes

1. 4 0. 9 1. 7 -0.7 3.5 0.3 1.8 0.5

-2.7 2.0 0.1 -0.2

1. 7 0.0 -0.7 -0.1 3.9 2.1 0.1 -0.7 0.1 1. 6 1. 2 -0.3

-0.4 -1.1 0.6 -3.5

1. 2 -3.0 1. 4 -0.8

-0.4 -0.6 -0.6 -0.4

-2.4 -3.9 0.4 -1.6

1. 2 -2.7

-0.6 Mean 0.007% 1 Std. Dev 1.70%

52 3- 6

, o Figure 3-1 Octant Assembly Pinvise Power 3.28 w/o U235,1.0 w/o Bcc 700 ppm, HFP 0 kid /MTU N-eff Max Pin PD0 1.11441 1.091 CASMO 1.11!.50 1.091 IT 1.043 1.009 1.047 1.012

.996 .976 BP

.995 .974

.984 .971 .956 .951

.984 .976 .954 .958

.982 .971 .956 .949 BP

.9'83 .975 .953 .947

.987 .972 * *

• BP

.987 .969 .959 .970 .997 1.003 .994 .987 .994 1.009 1.022 1.040 1.002 .997 .981 .996 1.008 1.020 1.036 1.032 1.030 1.029 1.033 1.041 1.051 1.066 1.091* PDQ l.032 1.033 1.032 1.036 1.043 1.052 1.065 1.091* CASMO S2 3- 7

Figure 3-2 Octant Asse=bly Finwise Power

. 3.28 w/o U235, 1.0 w/o B 4C 700 ppm, HFP 5000 MWD /MTU K-eff Max Pin PD0 1.10905 1.046 CASMO 1.10916 1.054 IT 1.035 1.015 1.036 1.012 0.996 0.996 0.992 0.996 BP 0.984 0.986 0.989 0.989 0.981 0.984 0.989 0.988 0.981 0.985 0.989 0.992 0.979 0.983 0.989 0.993 BP 0.984 0.989 BP 0.994 0.993 0.993 0.981 0.989 0.995 0.994 0.992 0.991 0.994 0.997 0.996 0.995 0.998 1.006 0.988 0.992 0.997 0.995 0.993 0.996 1.004 1.010 1.011 1.012 1.012 1.013 1.017 1.026 1.046* PDQ 1.011 1.012 1.013 1.015 1.017 1.021 1.029 1.054* CASMO S2 3-8

Figure 3-3 l

l Octant Assembly Pinwise Powers l

l 3.28 w/o U235, 1.0 w/o B 4C

- 700 ppm, HFP 10,000 MWD /KIU K-eff Max Pin PD0 1.08656 1.027 CASMO 1.08709 1.029 IT 1.027* 1.017 1.029* 1.014 0.996 1.007 0.991 1.009 BP 0.984 0.995 1.006 1.010 0.981 0.990 1.009 1.006 0.982 0.993 1.007 1.015 0.978 0.988 1.009 1.019 BP 0.984 0.998 1.012 1.004 0.992 0.980 1.001 BP 1.016 1.007 0.989 0.985 0.994 1.003 0.998 0.989 0.986 0.988 0.982 0.990 1.006 0.994 0.985 0.983 0.986 0.997 0.999 1.002 1.001 0.998 0.998 1.004 1.019 PDQ 0.998 1.000 1.003 1.002 1.G90 1.001 1.006 1.027 CASMO S2 3-9

l-l Figbre 3-4A l

Octant Assembly Pinwise Powers 3.28 w/o U235, 1.0 w/o B 4C 700 ppm, HFP i

20,000 MRD/MTU K-eff Max Pin PDO 1.01674 1.021 CASMO 1.01830 1.027 IT 1.017 1.014 1.020 1.014 0.998 1.010 0.992 1.014 BP 0.9 89 1.000 1.013 1.017 0.984 0.995 1.016 1.014 0.987 0.998 1.013 1.021*

0.982 0:.994 1.016 1.027* BP 0.988 1.003 BP 1.016 1.007 0.994 0.982 1.006 1.023 1.011 0.990 0.986 0.995 1.004 0.999 0.989 0.984 0.984 0.981 0.992 1.009 0.995 0.983 0.978 0.979 0.992 0.994 0.997 0.995 0.992 0.990 0.993 1.003 PDQ 0.994 0.995 0.998 0.996 0.993 0.992 0.995 1.010 CASMO S2 3-10A Rev. 1

l Figure 3-4 B .

Octant Assembly Pinwise Powers

. 3.28 w/o U235, BPR Removed 700 ppm, HFP 20,000 MWD /MTU K-eff Max Pin PD0 1.02862 1.070 CASMO 1.02764 1.074 IT 1.005 1.016 1.0 10 1.012 0.993 1.028 0.989 1.035 GT 0.984 1.013 1.047 1.058 0.981 1.003 1.052 1.045 0.981 '1.011 1.049 1.070*

0.978 1.001 1.053 1.074* GT 0.981 1.0 19 1.055 1.029 0.993 0.978 1.026 GT 1.060 1.036 0.985 0.971 0.993 1.016 1.002 0.976 0.960 0.951 0.968 0.986 1.024 0.993 0.970 0.955 0.947 l

l 0.966 0.972 0.978 0.974 0.964 0.957 0.955 0.962 PDQ O.972 0.975 0.980 0.975 0.967 0.960 0.958 0.971 CASM0 S2 3-10 B Rev. 1

Figure 3-5 Octant Assembly Pinwise Powers

- 3.28 w/o U235 BPR Removed 700 pps, HFP 40,000 MWD /MTU .

K-eff Max Pin PDQ 0.88438 1.024 CASMO - 0.88538 1.038 IT 1.002 1.006 1.007 1.013 1.001 1.011 0.995 1.018 GT 0.999 1.008 1.019 1.023 0.992 1.006 1.026 1.028 0.998 1.007 1.0 19 1.024*

0.991 1.004 1.026 1.038* GT 0.996 1.008 1.018 1.010 0.997 0.989 1.013 GT 1.031 1.0 16 0.994 0.991 0.998 1.005 1.001 0.992 0.984 0.979 0.983 0.995 1.010 0.998 0.983 0.976 0.971 0.987 0.989 0.991 0.990 0.985 0.981 0.978 0.979 PDQ 0.988 0.989 0.991 0.989 0.985 0.981 0.979 0.985 CASMO S2 3-11

4.0 POWER DISTRIBUTION COMPARISCNS 4.1 EPRI-NODE-P Model The three-dimensional nuclear code used at Duke Power for reload design is EPRI-NODE-P. This code is used for all maneuvering analyses, core follow, and physics test data predictions where three-dimensional core power distributions are required. In this section, comparisons of measured and EPRI-N0DE-P calculated values will be shown for both radial and total peak powers.

Comparisons were performed on a total of 62 reactor state points covering 1

Cycles 7 and 8 of Oconee Unit 1 The Oconee core was modeled using quarter core symmetry. Each fuel assembly was modeled with one radial and 16 equidistant axial nodes.

-The active stack height was set at 139 inches, the approximate densified fuel height. Control rods could be positioned continuously in this model, with a maximum inserted length to 5 inches above the bottom of the fuel stack as in the reactor itself.

Methods have been developed at Duke Power to generate fuel lattice (and lattice regions for PDQ) neutronics characteristics using CASMO. The data generated by CASMO is prepared by linking cod - for use in EPRI-N0DE-P and PDQ. Figure 4-1 shows a flow chart of the general methodology employed. In generating the input for CASMO, EPRI-NODE-P, and PDQ for this topical, these methods were applied exactly as for a reload design.

52 4-1

i EPRI-N0DE-P uses fuel km fits versus moderator temperature for rodded and l

unrodded conditions. Assemblies having burnable poison rods are treated f similarly. All km fits are referenced to either moderator temperature or fuel exposure. Also, fuel temperature reactivity changes are included. Therefore, EPRI-NODE-P explicitly models the effects of Moderator and Doppler feedbacks.

Since EPRI-NODE-P does not explicitly account for the core reflector and baffle, the assembly radial powers from EPRI-NODE-P were normalized to discrete pin model PDQ07 radial powers. A single normalization was performed when the core reached equilibrium xenon and samarium conditions and burnable poison depletion was well underway (~100 EFPD).

4.2 Oconee Fuel Cycle Simulations Using the EPRI-N0DE-P model described in Section 4-1, cycles 7 and 8 of Oconee Unit I were depleted in a core follow mode. At each statepoint, EPRI-N0DE-P calculates the critical boron concentration (the boron search option yields boron to il PPM so no correction is applied to the calculated critical boron),

the 3-D power distribution, and the axial offset among other things. Tables 4-1 and 4-2 show the statepoints, conditions, measurements, and predictions-for cycle 7 and 8 respectively.

Assembly radial and total powers at BOC, MOC, and EOC-7 are shown in Figures 4-2 through 4-7. Several representative axial power distributions for BOC, MOC, and EOC 7 are shown in Figures 4-8 through 4-13. The highest power assembly and the APSR assembly, where the part length rod causes severe power changes along the assembly, are illustrated.

S2 4-2

r The same type of data from EPRI-NODE-P and measurement for cycle 8 is shown in Figures 4-14 thru 4-19 (radial and total powers) and 4-20 thru 4-25 (axial

< powers). Here the axial powers shown are from the highest power assembly and the central gadolinium loaded test assembly.

Additional information on the design of these cycles is given in the respective reload reports (References 3 and 4).

4.3 PDQ07 Model Radial power distributions from PDQ07 depletions of Oconee 1 Cycles 7 and 8 are shown in Figures 4-26 thru 4-31 for BOC, M0C, and EOC. This 2-D model uses 2 group cross sections from CASMO and includes moderator and fuel temperature feedback effects on the fuel cross sections. (See section 3.1.1.2 for further discussion of the PDQ model.) All power distributions shown were from the hot full power, all rods out (except APSR's, which are modeled as control rods partly inserted) depletion.

4.4 Conclusions EPRI-NODE-P from a CASMO data base is shown to give consistently good power distributions when compared to measured power distributions. The assembly radial, total, and axial powers are demonstrated to be acceptable from a qualitative point of view for the data shown in this section. The quantitative statistical results are shown in Section 5.

52 4-3

Table 4-1 Oconee Unit 1, Cycle 7 Statepoints Critical Boron Power Rod Grocos % WD PPM Axial Offset, %

Point # EFPD  % FP Group 7 Group 8 (Meas / Calc) Meas / Calc) 6.9 98.9 90.0 28.5 1140/1124 -0.47/-5.42

/-4.52 2 10.3 99.5 92.7 29.0 /1121 3 14.8 99.7 93.6 29.5 1130/1115 3.27/-4.48 4 19.8 99.8 91.2 27.0 1060/1103 2.33/-3.59 5 36.0 99.7 93.0 29.0 1040/1074 -4.23/-4.33 6 48.7 99.7 94.2 29.0 1020/1048 -3.46/-3.85 7 53.7 99.8 94.5 29.3 1010/1037 -4.87/-4.05 8 66.9 96.3 94.0 29.4 988/1016 -4.64/-3.74 9 64.6 100.1 95.2 29.1 978/991 -4.29/-3.70 10 84.5 99.9 93.5 28.9 944/968 -5.01/-3.92 11 93.5 99.9 94.7 29.3 866/948 -6.81/-3.97 12 107.8 100.0 95.5 29.5 890/914 -4.77/-3.95 13 119.4 100.0 94.9 29.2 863/885 -5.17/-3.91 14 142.8 100.0 95.0 29.0 809/826 -3.93/-3.70 15 160.3 100.0 95.2 29.1 770/781 -4.30/-3.65 16 171.3 100.0 95.3 29.0 744/752 -3.64/-3.57 17 178.3 100.0 95.9 29.1 705/734 -4.47/-3.49 18 191.6 99.5 94.8 2'9.1 685/699 -3.69/-3.69 19 209.0 99.9 95.3 29.1 618/651 -4.73/-3.59 20 224.0 99.7 94.8 29.1 580/610 -3.30/-3.67 21 238.9 100.0 94.0 29.1 542/567 -2.00/-3.87 22 254.8 100.0 94.9 29.2 494/523 -2.50/-3.78 23 271.0 100.0 94.7 27.4 449/478 -1.44/-2.34 24 285.1 100.0 95.2 27.5 411/439 -1.44/-2.45 25 300.1 100.0 95.9 27.4 365/397 -1.34/-2.35 26 312.9 100.0 93.8 28.0 332/358 -2.28/-3.49 27 327.3 100.0 95.6 28.0 247/319 -0.88/-3.00 28 342.8 100.0 94.3 28.2 247/274 -1.55/-3.61 29 357.8 100.0 95.9 29.0 202/232 -0.78/-3.77 30 372.8 100.0 95.6 29.1 159/190 0.27/-3.89 31 387.8 100.0 95.1 29.9 118/147 0.34/-4.57 32 399.8 100.0 95.7 29.9 89/113 1.35/-4.26 33 402.9 71.4 84.1 25.4 105/156 -0.23/-1.86 S2 4-4

Table 4-2 i

l Oconee Unit 1, Cycle 8 Statepoints Critical Baron i

Rod Groups, % WD PPM Axial Offset, %

Power Point # EFPD  % FP Group 7 Group 8 (Meas / Calc) (Meas / Calc) l 1 4.0 100.0 90.3 25.9 1176/1053 0.17/-4.65 2 7.0 99.9 89.3 25.9 1030/1048 0.42/-5.03 3 9. 0 98.5 88.9 25.9 1035/1047 -1.01/-5.12 4 21.0 100.0 93.6 25.6 991/1030 -0.27/-2.69 5 37.9 100.0 93.8 25.5 964/998 -0.57/-2.68 6 52.9 100.0 93.4 25.1 949/966 -0.53/-2.64 7 68.9 100.0 92.3 25.1 926/930 -1.67/-3.34 8 83.0 100.0 94.1 25.1 899/901 -1.54/-2.58 9 99.1 100.0 94.7 25.6 868/864 -1.52/-2.89 10 113.8 100.0 95.3 25.6 835/830 -1.60/-2.83 11 129.8 100.0 95.2 25.6 794/791 -1.79/-2.94 12 145.7 100.0 95.0 25.7 760/752 -2.21/-3.12 13 160.7 100.0 95.2 26.2 724/715 -2.45/-3.41 14 172.8 100.0 95.4 26.2 659/684 -2.31/-3.35 15 189.7 100.1 93.2 25.6 639/639 -2.43/-3.64 16 203.3 99.9 95.1 25.1 606/605 -3.20/-2.68 17 219.3 100.1 94.9 25.1 560/563 -2.53/-2.84 18 234.3 100.0 93.7 25.1 523/522 -2.59/-3.19 19 249.3 100.0 94.3 25.1 480/483 -2.49/-3.05 20 263.6 100.0 94.9 25.9 452/444 -2.22/-3.39 21 278.4 100.0 95.0 24.2 394/405 -1.32/-2.33 22 294.3 96.1 93.1 25.3 351/368 -2.86/-3.01 23 309.3 100.1 95.6 25.3 310/320 -1.76/-2.92 24 324.3 100.0 96.1 25.2 268/280 -1.63/-2.74 25 340.2 100.1 95.0 25.1 218/235 -1.68/-3.06 26 355.2 99.9 93.5 24.6 176/193 -1.16/-3.20 27 370.3 100.1 95.2 25.2 137/153 -0.77/-3.06 28 385.2 100.0 95.0 25.0 94/113 -0.19/-3.00 29 400.3 100.0 95.0 25.0 48/72 -0.54/-3.03 52 4-5

Figure 4-1 NODE-P_and PDQ Setup Methodology from CASMO 3

CASMO Depletion and Branches of Assembly Compositions Lattice Data Cross Sections in for N00E-P 2 Groups for Fuel and Km, M2, k/v, I Nonfuel (Lumped Burnable oaxe,etc.as$, Poisons, Water Holes, etc.)

function of exposure ,

boron, temperature, etc.

V

~~Y' 2-0, Quarter Core P0Q with Normalization of Thermal Feedback. Radial NODE-P to PDQ Radial =: Assembly Average and Pin i Power Distribution Power Distributions throughout cycle.

'I 3-0, Quarter Core NODE-P with Thermal Feedback.

Radial and Axial Power Distributions, Critical i

Baron, Rod Worths, Reactivity Coefficients, etc., throughout cycle.

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5. STATISTICAL ANALYSIS 5.1 Observed Nuclear Reliability Factor Derivation This section determines quantitatively the statistics from the comparison of measured and calculated power distributions which are discussed in Section 4.

Normal distribution theory is used in deriving calculational uncertainties.

The derivation of the statistical method used to determine the Observed Nuclear Reliability Factors (0NRF) is given in Supplement 2, Section 5.1 of Reference

1. The ONRF is ultimately defined by the relationship:

ONRF = N - 6 + K

• S(0) (5-1)

A Where 6 = f - N = (E 09 ) + n (5-2) i=1 04=C9 -M 9 the th 9

difference, 1 1 i i N (5-3)

Cg is the 4th calculated value (radial or peak)

Mg is the 9th measured value (radial or peak) n f = (I c;) + n (5-4) i=1 n

N = (I M,) + n (5-5) i=1 K is the one-sided 95/95 tolerance factor S(0) is the standard deviation of the differences 52 5-1

The ONRF from equation 5-1 will be used as a multiplicative factor applied to EPRI-NODE-P calculated powers such that:

ONRF x C > M (5-6) for 95% of the population with a confidence factor of 95%. Separate ONRF's are determined for the assembly radial and assembly peak powers.

In practice, other uncertainty factors are combined with the ONRF to produce a statistically combined uncertainty factor (SCUF) for centerline fuel melt, DNB, and LOCA limiting power distribution limits.

5.2 Normality Test Results Using equation 5-2, the variable D was used to form a frequency distribution.

Histograms of these difference distributions are shown in Figures 5-1 to 5-6 for the radial and peak differences for cycle 7 and 8 and both cycles combined.

Each of these was analyzed for normality using the D' test from ANSI N15.15-1974 (Reference 5).

For the normality test of the difference distributions, the calculated (C) and measured (M) data were grouped as follows:

1) Reactor Cycle: 7 or 8

2) Combined Cycles: 7 and 8

3) Type: Assembly peak or assembly radial Using the engineering judgement that only RPDs greater than the core average are of concern, pairs of C and M where both are > 1.0 will be treated. Table 5-1 shows the results of the normality tests. With a level of significance of 52 5-2

5% the 2.5% and 97.5% percentage points define the critical values in Table 5 of reference 5. The D' values calculated must fall between the critical values to be deemed normal. Table 5-1 shows all ccmbinations to be normal except t!e assembly peak power differences for the combined cycles 7 and 8. In this case, the D' value falls just outside the lower critical value and will be classified as nearly normal. The histogram of this data, Figure 5-6, appears normal.

Therefore, the assumption of normality will be used for all data sets.

5.3 Observed Nuclear Reliability Factors for EPRI-N00E-P In this subsection the statistical treatment developed in Section 5.1 is l utilized to develop ONRFs for the CASMO method of reload design for Oconee Nuclear Station. The reliability factors are determined in a manner consistent with their use in reload design.

Reload design determines, among other things, the margin to centerline fuel melt, DNB, and LOCA limited kw/ft by comparing the maximum peak power in the core to a predetermined limiting value. Therefore, the ONRFs are determined by comparing the maximum peak power in the core at each statepoint listed in Tables 4-1 and 4-2. This is done for both the peak radial power and peak total power.

Table 5-2 shows the data used in the calculation of the ONRF as well as the ONRF itself. The combined radial ONRF is 1.038 and the total peak ONRF is 1.059.

52 5-3

5.4 Quantitative Comparisons of EPRI-N00E-P to Measurements By analyzing the variable 0 as defined in equation 5-2, the accuracy of EPRI-NODE-P can be assessed. Four important statistical properties of 0 are discussed.

D is the mean of the difference between EPRI-N00E-P and measured assembly powers. For cycles 7 and 8, 5 is -0.02727 for the radials and -0.02544 for the peaks (Table 5-2). The standard deviation of the differences indicates the spread of the values of 0 about 5. For cycles 7 and 8, S(0) is 0.01185 for the radials and 0.03318 for the peaks. .

The mean of the absolute differences ABS (0) and its standard deviation can be combined to give limits on this variable. 95% confidence limits on the means were given by:

ABS (0)U,L = ABS (0) t(.05,n) x S(ABS (0)) (5-7)

Vn For the combined radials, equation 5-7 gives:

ABS (0)U,L = 0.0273 0.0025 and for the peaks:

ABS (0)U,L = 0.0343 t 0.0050 Tables 5-3 and 5-4 present summary 0 statistics for peaks and radials where (C,M) > 1.0 for all pairs considered, as in the normality test in Section 5.2 52 5-4

5.5 ONRF Confirmation In order to confirm that the combined ONRF's calculated in Table 5-2 do indeed meet the 95% criteria, a comparison of measured and (calculated

• ONRF) is performed. For all of the 62 state points used in the ONRF calculation, the maximum measured radials never exceeded 1.038 x maximum calculated radials; and the maximum measured peaks never exceeded 1.059 x maximum calculated peaks.

Furthermore, of the 1024 radials where (C,M) > 1.0, 51 (or 4.38%) measured radials exceeded 1.038 x calculated radials; and of the 1233 peaks, 28 (or 2.27%) measured peaks exceeded 1.059 x calculated peaks.

5.6 Conclusions A statistical analysis of EPRI-NODE-P calculated power distributions, based on a CASMO physics data base, and plant measured power distributions has shown that the observed Nuclear Reliability Factors are:

ONRF Radial Powers = 1.038 CNRF Peak Powers = 1.059 These were determined from analysis and measurement of Oconee Unit 1 but are applicable to all 3 units for the following reasons:

1. All three units have identical incore detector systems.

2. All three units are manufactured by the same vendor and use similar fuel.

3. Calculations for all three units will be performed using the same methods and procedures.

52 5-5

Table 5-1 Difference Distributuion Normality Tests For (C, M > 1.0) - 5% Level of Significance Assembly Radial Power D' D' (P = 0.975) Remarks Cycle N D' (P = 0.025) 3027.4 3028.6 Normal 7 484 2971.2 3516.8 3567.0 Normal 8 540 3504.0 9263.3 9299.4 Normal 7,8 1024 9179.5 Assembly Peak Power D' D' (P = 0.975) Remarks Cycle N D' (P = 0.025) 38/8.5 3919.2 Normal 7 575 3851.8 4717.8 4795.3 Normal 8 658 4717.6 12098.8 12278.4 Nearly 7,8 1233 12138.4 Normal S2 5-6

Table 5-2 EPRI-NODE-P ONRF Calculation Corewise Maximum Radial

- Cycle N_ R 5 S(0) K(95/95) ONRF 7 33- 1.3592 -0.02741 0.01299 2.186 1.041 8 29 1.3586 -0.02711 0.01062 2.232 1.037 Combined 62 1.3589 -0.02727 0.01185 2.015 1.038 Corewise Maximum Peak Cycle N_ R 5 S(0) K(95/95) ONRF 7 33 1.5719 -0.03059 0.04138 2.186 1.077 8 29 1.5395 -0.01959 0.01936 2.232 1.041 Combined 62 1.5567 -0.02544 0.03318 2.015 1.059 Where, ONRF = - 5 + K

• S(0)

E and K is from Table 2.4 of reference 6.

d 52 5-7

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Table 5-3 Difference Means and Standard Deviations for Peaks ((C,M) >'1.0) f Cycle N 5 S(0) ABS (0) S(ABS (0))

, 7 575 0.016424 0.051415 0.044417 0.030618 8 658 0.009706 0.036473 0.029542 0.023464 7,8 1233 0.012839 0.044184 0.036479 0.028026 I

52 5-8

Table 5-4 Difference Means and Standard Deviations for Radials ((C.M) > 1.0) l Cycle N D S(D) A85(D) S(ABS (0))

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6.0 References

1. Duke Power Company Oconee Nuclear Station Reload Design Methodology NFS-1001A, April 1984.

2. Edenius, M. et al. "CASMO-Bencmark Report" STUDSVIK/RF-78/6293, March 1978.

3. Oconee Unit 1, Cycle 7 Reload Report, BAW-1660, March, 1981.

4. Oconee Unit 1, Cycle 8 Reload Report, BAW-1774, April, 1983.

5. American National Standards Institute, Inc., " Assessment of the Assumption of Normality (Employing Individual Observed Values)," ANSI NIS, 15-1974, 1974.

6. Owen, D. B. , " Factors For One-Sided Tolerance Limits And for Variables Sampling Plans," SCR-607, Sandia Corporation Monograph, March 1963.

52 6-1

L NRC QUESTIONS AND DUKE POWER COMPANY RESPONSES 1

O l

l l

l i

1. MQvq y

y kg UNITED STATES NUCLEAR REGULATORY COMMISSION l

g WASHINGTON, O. C. 20555 o, -

'% / June 4, 1985

• FoNetsNos. 50-269, 50-270 and 50-287 i

Mr. Hal B. Tucker Vice President - Nuclear Production Duke Power Company P. O. Box 33189 422 South Church Street Charlotte, North Carolina 28242

SUBJECT:

RELOAD DESIGN METHODOLOGY II REPORT - REQUEST FOR ADDITIONAL INFORMATION ,

Re: Oconee Nuclear Station, Units 1, 2 and 3 .

(

By letter dated April 3,1985, you submitted for our review, the "0conee Reload Design Med c'Mogy II" report, DPC-NE-1002. The report presents an alternate nuclear oesign code sequence to that currently approved in DPC-NFS-1001A. The ma,ior difference is that CASM0-2 replaces EPRI-CELL for generation of assembly average reactor physics properties and for generation of two neutron energy group cross sections. Changes have also been made to the thermal hydraulic section to include the use of the BWC heat flux correlation and to the mechanical fuel analysis section. In your letter, you stated that you are planning to use this revised methodology for the upcoming Oconee 1, Cycle 10 reload.

We have been reviewing the Oconee Reload Design Methodology II report, and have determined that we need additional information to complete our review.

To meet your schedule, we request that you respond to the enclosed list of s questions within 30 days of receipt of this letter. This request for infomation affects fewer than ten respondents; therefore, OMB clearance is not required.

Sincerely, o J.

Y Y a &r d ' i

/ John F. Stolz, Chief

@ erating Reactors Erahch #4 Division of Licensing

Enclosure:

As Stated cc w/ enclosure: '

See next page r

s Mr. H. B. Tucker Oconee Nuclear Station, Units Duke Power Company Nos. 1, 2 and 3 .-

l l cc: '

l Mr. William L. Porter c l Duke Power Company l

P. O. Box 33189 "

422 South Church Street Charlotte, North Carolina 28242 J. Michael McGarry, III, Esq.

Bishop, Liberman, Cook, Purcell & Reynolds 1200 Seventeenth Street, N.W. -

Washington, D.C. 20036 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division '

Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Manager, LIS I NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. J. C. Bryant Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission, Region II 101 Marietta Street, N.W.

Suite 3100 Atlanta, Georgia 30303 Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control .

2600 Bull Street Columbia, South Carolina 29201 Office of Intergovernmental Relations 116 West Jones Street Raleigh, North Carolina 27603 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621

Et: CLOSURE P

REOUEST FOR ADDITIONAL INFORMATION DUKE POWER COMPAf!Y OCONEE NUCLEAR STATION RELOAD DESIGN METHODOLOGY

1. Pg. 1-1. What CASM0 feature (s) constitute an improvement of the methodology over EPRI-CELL?. <

2. Pg. 4-3. Was there ever any information of consequence during reload review which was not documented?

(I

3. Pgs. 51 2-1, 51 2-3. Why were critical baron concentrations calculated /

with all rods out except group 8 only? Why not the other banks? j

4. Pg. S1 4-1. If the ejected rod worth is defined in terms of measured worth why did you stop measuring? Shouldn't the definition be changed f^

to calculated worth? l

5. Pg. 51 4-1. Since the ejected rod worth is no longer measured, why is paragraph 4.1 needed? The calculation description in paragraph 4.2 seems adequate.

6. Pg. 51 4-3 4.3. One measurement is hardly worth mentioning as a comparison for a measurement which is not being used and is known to be difficult and inaccurate.

7. Pg. S1 4-3. Summary. The comparison cannot be used to support the conclusion. ,

8. Pg. 51 5-3. Table 5-1. Only two measurements? What is 'the statistical significance of the standard deviation?

9. Pg. S2 3-1 3.1. What is the Metal / Water for the Oconee vs the Kritz criticals.

10. Pg. 52 3-4. What are the differences of the CASM0 referred to up to now and CASMO-2E (this difference is in addition to the Oconee-Kritz differences).

11. Pg. 52 3-6. Can the 14x14 and 15x15 assemblies be used together? Their difference in their mean looks more as a systematic (bias) deviation due to the presence of the absorbers. the significance of the combined mean value is; questionfole. Have the similarities and differences in M/W, pitch;-diameter etc. been looked at?

12. Pg. 52 3-10. Why is the 20,000 MWD /MTU, 700 ppm and burnable poison, rods in, case missing? ,

l l

1 DuxE POWER COMPANY P.O. DOx 33189 CHARLOTTE. N.C. 28242 HAL B. TL*CKER Truenon esce ,.....en (704) 373-4 531

..u . .-evio, June 12, 1985 Mr. Harold R. Denton, Director Office of Nuclear Re:tter Regulation U. S. Nuclear Regulatory Commission Washington, D. C.~ 20555 Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4

Subject:

Oconee Nuclear Station Docket Nos. 50-269 -270, -287

Dear Sir:

By a letter dated June 4,1985, the NRC for= ally transmitted a list of questions, comprising a request for additional information, concerning a Duke Power :echnical report, "Oconee Reload Design Methodology II",

DPC-NE-1002, which was submitted for NRC review by an Aptil 3, 1985 letter. The responses are provided in Attachment 1, and the revised pages for DPC-NE-1002 (which are associated with the responses in Attachment 1) are included as Attachment 2.

As noted in the June 4th letter, Duke plans to use the revised method-ology in the upcoming Oconee 1, Cycle 10 reload analysis. A timely consiceration of the attached information, in order to co=plete the review, would be appreciated.

Very truly yours, 2 W Hal B. Tucker RFH: sib Attachment cc: Dr. J. Nelson Grace, Regional Administrator Mr. J. C. Bryant U. S. Nuclear Regulatory Commission NRC Resident Inspector Region II Oconee Nuclear Station 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Ms. Helen Nicolaras Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Mr. Harold R. Denton, Director June 12, 1985 Page Two bec: P. M. Abraham K. S. Canady R. H. Clark J. L. Eller T. C. Geer ,

S. G. Godwin P. F. Guill R. M. Gribble M. A. Haghi G. P. Horne N. A. Rutherford R. G. Snipes G. B. Swindlehurst R. T. Bond ONS J. W. Collier ONS >

T. C. Matthews ONS Group File: OS-801.01 2

o ATTACIDENT 1 Responses to Requests for Additional Infor=ation 4

Duke Power Co=pany Oconee Nuclear Station Response To Request For Additional Information Duke Power Company Oconee Nuclear Station Reload Design Methodology Request: 1. Pg. 1-1. What CASMO feature (s) constitute an improve-

=ent of the methodology over EPRI-CELL?.

Response: 1. CASMO is a more user friendly code and is less expensive than EPRI-CELL and Color Set PDQ calculations. The same CASMO execution can be used to define cross sections for the various fuel assembly components such as fuel pins, burnable poisons, control rod guide tubes, etc. Also, since CASMO is an assembly transport theory code it is calculationally superior to Color Set PDQ for determining assembly-wise physics parameter for input to nodal codes.

Request: 2. Pg. 4-3. Was there ever any information of consequence during reload review which was not documented?

Response: 2. No, the wording on page 4-3 has been modified for clarity in Revision 1.

Request: 3. Pgs. S1 2-1, S1 2-3. Why were critical boren concentrations calculated with all rods out except group 8 only? Why not the other banks?

Response: 3. Recent cycles at Oconee have been operated in an all rods out configuration (feed and bleed) except for the part length group 8 rods which are used for power-imbalance control. For this reason nearly all the measured data available for comparisons is at all rods out conditions.

Table 3-2 on page S1 3-5 may be used to infer the accuracy of CASMO at predicting rods in boron concentrations since it shows the calculated and measured worth of control rod groups 5-7 at HZP, BOC conditions.

Request: 4. Pg. S1 4-1. If the ejected rod worth is defined in terms of measured worth why did you stop measuring? Shouldn't -

the definition be changed to calculated worth?

5. Pg. S1 4-1. Since the ejected rod worth is no longer measured, why is paragraph 4.1 needed? The calculation description in paragraph 4.2 seems adequate.

6. Pg. S1 4-3 4.3. One measurement is hardly worth mentioning as a comparison for a measurement which is not being used and is known to be difficult and inaccurate.

7. Pg. S1 4-3. Sum =ary. The comparison cannot be used to support the conclusion.

Response: 4- 7. Since the ejected rod worth is not measured any longer questions 4-7 have been answered by deleting chapter 4 of supplement 1. This modift:ation is reflected on the revised pages of the report.

Request: 8. Pg. S1 5-3. Table 5-1. Only two measurements? What is the statistical significance of the standard deviation?

Response: 8. Question 8 is answered by modifying pages S15-2 and S1 5-3 to remove any reference to the mean and standard deviation.'

Request: 9. Pg. S2 3-1 3.1. What is the Metal / Water for the Oconee vs. the Kritz criticals.

Response: 9. The metal to water ratio for Oconee is given in the original Oconee reload design methodology, NFS-1001A and is 0.828227. The Kritz critical experiments were performed on lattice configurations similar to standard PWR geometries. However, the actual dimensions are proprietary and have not been published.

Request: 10. Pg. S2 3-4. What are the differences of the CASMO referred to up to now and CASMO-2E (this difference is in addition to the Oconee-Kritz differences).

11. Pg. S2 3-6. Can the 14x14 and 15x15 assemblies be used together? Their difference in their mean looks more as a systematic (bias) deviation due to the presence of the absorbers. The significance of the combined mean value is questionable. Have the similarities and differences in M/W, pitch, diameter etc. been looked at?

Response: la-ll. Questions 10 through 11 are satisfied by rewriting para-graph 3.3 on page S2 3-4 and modifying page 32 3-6. These modifications are included in the revised report pages.

The NRC has previously approved the combination of the two Kritz experiments in the Northern States Power topical report entitled " Qualification of Reactor Physics Methods For Application to PI Units," NSPNAD-810lP. December 1,1981.

Request: 12. Pg. S2 3-10. Why is the 20,000 MWD /MTU, 700 ppm and burnable poison, rods in, case missing?

Response: 12. The figure in question has been provided in the revised report pages.

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ATTACICENT 2 Revision 1 to DPC-NE-1002

OPC - NE - 1002 PAGES TO BE REVISED l

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Pellet 0.D. dimensions are used to calculate cladding strain because the strain itself is caused by pellet thermal expansion. There are three major conditions in this calculation that make it conservative. The first is the extreme power change that is used to simulate the worst case peaking. The second is that the pellet is assumed to be in hard contact at inititation of the ramp. This is a conservative assumption since the power ramp is initiated from a very low power level and pellet / cladding contact is not expected to occur at this low linear heat rate. The third conservatism is that the pellet is non-compliant and that all of the pellet thermal expansion results directly in cladding strain.

4.4 Claddino Stress Analysis The methodology of Reference 1 has been revised. The current methodology is consistent with that of References 5 and 12.

1 The static stress analysis uses design stress intensity limits on mechanical properties based on the requirements of ASME Code Article III-2000. Thus, the design stress intensity value for Zircaloy-4 is the lowest of the following:

(1) one-third of the specified minimum tensile strength at room temperature (2) one third of the tensile strength at temperature (3) two-thirds of the specified minimum yield strength at room temperature (4) two-thirds of the yield strength at temperature 4-3

10. J. F. Stolz (N".C) to R. J. Rodriguez (SMUD), letter " Rancho Seco Nuclear Generating Station - Evaluation of Mark-BZ Fuel Assembly Design", November 16, 1984.

11. H. B. Tucker to J. F. Stolz, Oconee Nuclear Station Docket Nos. 50-269,

-270, -287, March 8, 1985.

12. Oconee Unit 3, Cycle 9 Reload Report DPC-RD-2005 June 1985. 1

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I 10-2

TABLE OF CONTENTS

1. Introduction 51 1-1

2. Critical Baron Concentrations S1 2-1 2.1 Measurement Technique 51 2-1

2. 2 Calculational Technique 51 2-1 2.3 Comparisons of Calculated and Measured Results S1 2-2 2.4 Summary 51 2-3

3. Control Rod Worths 51 3-1 3.1 Measurement Techniques S1 3-1 3.2 Calculational Techniques 51 3-1 3.3 Comparisons of Calculated and Measured Results S1'3-2 3.4 Summary 51 3-3

4. Ejected Rod Worths 51 4-1 1

5. Isothermal Temperature Coefficients S1 5-1 5.1 Measurement Technique S1 5-1 5.2 Jalculational Technique S1 5-1 5.3 Comparison of Calculated and Measured Results 51 5-2 5.4 Summary 51 5-2

6. References S1 6-1 51-iii

i LIST OF TABLES l

Pace 2-1 Oconee 1 Cycles 7 and 8 BOC Critical Boron Concentrations at Hot Zero Power S1 2-4 2-2 Oconee 1 Cycles 7 and 8 Hot Full Power Critical Boron Concentrations from EPRI-NODE-P vs. Measured 51 2-5 2-3 Oconee 1 Cycles 7 and 8 Hot Full Power Critical Boron Concentrations from Quarter Core, Fine Mesh PDQ vs. Measured 51 2-7 3-1 Oconee 1 Cycles 7 and 8 Control Rod Worths At Hot Zero Power, BOC In Terms of Reactivity 51 3-4 3-2 Oconee 1 Cycles 7 and 8 Control Rod Worths At HZP, BOC In Terms of Boron 51 3-5 1

5-1 Oconee 1 Cycles 7 and 8 Isothermal Temperature Coefficients at HZP, BOC 51 5-3 51-iv

E

4. Ejected Rod Worths b

- This section deleted -

9 51 4-1 REV 1

5.3 Comoarison of Calculated and Measured Results A comparison of calculated and measured isothermal temperature coefficients at HZP and 80C for Oconee Unit 1, Cycles 7 and 8 is presented in Table 5-1. The agreement between these calculated and measured coefficients is very good; all

~

values are well within 0.3 x 10 4 ap/ F acceptance criteria.

t 5.4 Summary The comparison between calculated and measured isothermal temperature coef-ficients indicates that the CASMO based EPRI-NODE-P is an adequate predictor of isothermal temperature coefficients.

51 5-2

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Table 5-1 Oconee 1 Cycles 7 and 8 Isothermal Temperature Coefficients at HZP, BOC Boron Conc, Temp. Coeff, 10 4 Ap/ F Difference Cycle PPM Calculated Measured 10 4 ao/ F 7 1610 +0.229 +0.273 +0.044 8 1635 +0.08 +0.11 +0.03 1

5 51 5-3

3.3 Comoarisons of CASMO to Criticals An earlier version of CASMO has been benchmarked against the hot criticals at the Kritz facility. The version of CASMO used at Duke Power has improved flexibility and has certain edit capabilities not found in the early version.

The transport theory caiculations are the same in each version and hence the benchmarking discussed in Reference 2 applies to the CASMO used at Duke Power.

The results of the two PWR assembly experiments are given in Table 3-2. These experiments were performed with lattice configurations and dimensions which are similar to standard PWR assemblies. The actual dimensions are proprietary and have not been published.

The results of the two PWR experiments are combined to better represent assembly configurations present in Oconee cores. Current cycles contain assemblies which contain lumped burnable poisons as well as assemblies without burnable poisons. The Kritz experiment with burnable poisons was performed at less than 100 PPMB and the other experiment was performed at approximately 1000 PPMB. This range of baron concentrations is representative of Oconee fuel cycles. The results are given in Table 3-2 and give a CASMO to experiment deviation of 1.7%.

1 S2 3-4A REV 1

3.4- Conclusions The bounding PDQ to CASMO pin power difference and the CASMO to experiment error can be statistically combined to give the expected radial-local 4

uncertainty factor for CASMO based reload design. This factor is calculated I as:

+ ,0172 = 0.022 FR-L = 4 0142 This 2.2% conservatively accounts for pin power error t' rom the quarter-core PDQ07 model.

S2 3-4B REV 1

Table 3-2 l

CASMO vs. Experiment - Krit: Criticals '

Fission Rate Deviations L

r Deviation = 100 * (CASMO-EXPERIMENT)/ EXPERIMENT 15x15 PWR M0 2 14x14 PWR M0 2 with with Absorbers and Water Holes Water Holes 1.4 0. 9 1.7 3.5 0.3

-0.7 1.8 0.5

-2.7 2.0 0.1

1. 7 0.0 -0.2

-0.7 -0.1 3.9 2.1 0.1 0.1 1.6

-0.7

-0.4 1. 2 -0.3

- 1.1 0. 6 -3.5 1.2 -3.0 1.4

-0.4 -0.6 -0.8

-0.6 -0.4-

-2.4 -3.9 0.4 -1.6

1. 2 -2.7

-0.6 Mean 0.007% i Std. Dev 1.70%

e 52 3-6 4

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Figure 3-4A Octant Assembly Pinvise Powers 4

3.28 w/o U235, 1.0 w/o B 4C 700 ppm, HFP 20,000 M:1D/MIU K-eff Max Pin PD0 1.01674 1.021 CASMO 1.01830 1.027 IT 1.017 1.014 1.020 1.014 0.998 1.010 0.992 1.014 BP 0.989 1.000 1.013 1.017 0.984 0.995 1.016 1.014 0.987 0.998 1.013 1.021*

0.982 0.994 1.016 1.027* BP 0.988 1.003 BP 1.016 1.007 0.994 0.982 1.006 1.023 1.011 0.990 0.986 0.995 1.004 0.999 0.989 0.984 0.984 0.981 0.992 1.009 0.995 0.983 0.978 0.979 i-0.992 0.994 0.997 0.995 0.992 0.990 0.994 0.993 1.003 PDQ 0.995 0.998 0.996 0.993 0.992 0.995 1.010 CASMO S2 3-10A Rev. 1

1 Figure 3-4 B Octant Assembly Pinvise Powers 3.28 w/o U235, BPR Removed 700 ppm, HFP 20,000 MWD /MTU -

K-eff _

Max Pin Pno 1.02862 1.070 CASMO 1.02764 1.074 IT 1.005 1.016 1.010 1.012 0.993 1.028 0.989 1.035 GT l

0.984 1.013 1.047 1.058

0.981 1.003 1.052 1.045 l

0.981 '1.011 1.049 1.070*

0.978 1.001 1.053 1.074* GT "-

0.981 1.019 1.055 1.029 0.978 0.993 1.026 GT 1.060 1.036 0.985 0.971 0.993 1.016 1.002 0.976 0.968 0.986 0.960 0.951 1.024 0.993 0.970 0.955 0.947 0.966 0.972 0.9 78 0.974

• 0.9o4 0.957 0.955 0.962 0.972 0.9 75 0.980 0.975 PDQ 0.967 0.960 0.958 0.971 CASMO S2 3-10 B Rev.1