ML100250858

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Holtec Report, No. HI-2043321, Rev. 6, Critically Safety Analyses of BWR Fuel Without Credit for Boraflex in the Racks at the Harris Nuclear Power Station.
ML100250858
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Site: Harris Duke Energy icon.png
Issue date: 01/18/2010
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Holtec
To:
Office of Nuclear Reactor Regulation, Progress Energy Carolinas
References
HNP-09-087, TAC ME0012 HI-2043321, Rev 6
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Enclosure 1 to SERIAL: ,HNP-09-087 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 SECOND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ATTACHMENT 4 CRITICALITY SAFETY ANALYSES OF BWR FUEL WITHOUT CREDIT FOR BORAFLEX IN THE RACKS AT THE HARRIS NUCLEAR POWER STATION Holtec Report No. HI-2043321, Revision 6 (Non-Proprietary)

(34 Pages)

Eu.' I.- Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 HOLT EC INTERN AT I ON AL Fax (856) 797 - 0909 CRITICALITY SAFETY ANALYSES OF BWR FUEL WITHOUT CREDIT FOR BORAFLEX IN THE RACKS AT THE HARRIS NUCLEAR POWER STATION FOR PROGRESS ENERGY Holtec Report No: HI-2043321 Holtec Project No: 1430 Report Class: SAFETY RELATED

Summary of Revisions Revision 0:

Original issue.

Revision 1:

Client Revised GE 13 Fuel Assembly Design and Axial Segment Description.

Revision 2:

Incorporated client comments; expanded discussion on axial burnup distribution and selection of the reference assembly.

Revision 3:

Revised Appendix A to remove the Propriety Information note in the footer.

Revision 4:

Revised to increase the burnup requirements for enrichments of 1.5, 2.0 and 2.5 wt% to reduce the maximum keff below 0.99. The enrichment tolerance is changed to M . The polynomial burnup versus enrichment curve is recalculated as a linear fit. Figure 2 is replaced.

Revision 5 Revised to incorporate additional calculations to support NRC acceptance review questions.

Revision 6 Revised to incorporate new curves and tables for 4 and 7 year cooling times. Revision includes changes made to HI-2043306R6, R7 and R8.

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Table of Contents 1.0 PU RP O S E ........................................................................................................................... 1 2.0 ANALYSIS CRITERIA AND ASSUMPTIONS .......................................................... 3 3.0 ACCEPTANCE CRITERIA .......................................................................................... 4 4.0 DESIGN AND INPUT DATA ...................................................................................... 5 4.1 FUEL ASSEMBLY DESIGN ........................................................................................ 5 4.2 CORE OPERATION AND CONTROL RODS ............................................................ 5 4.3 STORAGE RACK DESIGN .......................................................................................... 6 5.0 M ETH O D O LO G Y ...................................................................................................... 7 5.1 GENERAL DESCRIPTION ........................................................................................... 7 5.2 AXIAL BURNUP DISTRIBUTION ............................................................................. 7 6.0 A NA LY SIS RESU LTS .................................................................................................. 9 6.1 EVALUATION OF TOLERANCE UNCERTAINTIES ............................................... 9 6.1.1 UNCERTAINTY IN DEPLETION CALCULATIONS ............................................... 10 6.2 ABNORMAL AND ACCIDENT CONDITIONS ....................................................... 10 6.2.1 ECCENTRIC LOCATION OF FUEL ASSEMBLIES .................................................... 11 6.2.2 TEMPERATURE AND VOID EFFECTS ....................................................................... 11 6.2.3 MIS-LOADED FUEL ASSEMBLY ACCIDENT ......................................................... II 7.0 C O N C L U SIO N S............................................................................................................... 12 8.0 RE FE REN C ES ................................................................................................................. 13 APPENDIX A BENCHMARK CALCULATIONS (Total Pages 27)

List of Tables Table 1: Maximum Reactivity at Various Initial Enrichment - Minimum Burnup Combinations (4 years cooling)

Table la: Maximum Reactivity at Various Initial Enrichment - Minimum Burnup Combinations (7 years cooling)

Table 2: BWR Fuel Characteristics Table 2a: Brunswick Core Operating Parameters for Depletion Table 2b: BWR Control Rod Characteristics Table 3: Tolerances and Their Reactivity Effects Report: HI-2043321 i Project 1430 Proprietary Information Denoted by Shaded Text

Table 4: Summary of Abnormal/Accident Conditions Table 5: BWR Boraflex Racks Burnup Versus Enrichment Requirement Table 6: Average Segment Axial Burnup Distribution Table 7: Assembly Design Reactivities (4 years)

Table 7a: Assembly Design Reactivities (7 years)

Table 8: Reactivity Effect of Temperature and Void Content 4.0 wt%, 45.0 MWD/kgU List of Figures Figure 1: Axial Burnup Distribution Figure 2: Minimum Fuel Burnup for Acceptable Storage of Spent Fuel of Various Initial Planar Average Enrichments (4 years)

Figure 2a: Minimum Fuel Burnup for Acceptable Storage of Spent Fuel of Various Initial Planar Average Enrichments (7 years)

Figure 3: Cross-Section of Typical Storage Cell (Calculational Model)

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1.0 PURPOSE The purpose of the present evaluation is to document the criticality safety of the BWR fuel storage racks in the spent fuel pools of the Harris Plant. The pool criticality analyses are performed under the very conservative assumption of the complete loss of Boraflex in the BWR storage rack. With the assumed loss of all Boraflex material, the temperature coefficient of reactivity is positive. Therefore the limiting calculations assumed a temperature of 150 OF, which is an administrative limit for the spent fuel pool. Higher temperatures are considered accident conditions for which the soluble boron normally present in the pool water would assure the reactivity is maintained below the regulatory limit of a k~ff of less than 0.95. Under normal storage conditions, partial credit is taken for the soluble boron in the pool water, credit for fuel burnup is taken, and credit for spent fuel cooling time is also taken. The criticality analyses use the MCNP4a code, a Monte Carlo code developed by the Los Alamos National Laboratory, with explicit modeling of actinide and fission product nuclide concentrations. CASMO4 was used for calculation of manufacturing tolerances, and to determine the burnup dependent nuclide inventories used in the MCNP4a calculations at the various burnups.

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0012 with an uncertainty of

+/- 0.0090 for MCNP4a, evaluated with a 95% probability at the 95% confidence level('). The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

Benchmark calculations, presented in Reference 6, indicate a negative bias and bias uncertainty of + 0.0025 for CASMO-4 evaluated with a 95% probability at the 95% confidence level(').

Since CASMO-4 is used to determine reactivity differences, the bias does not need to be applied to the results of the calculations. However, the bias uncertainty is included with the other uncertainties when determining the maximum ktff values.

As described in the USNRC guidelines, parametric evaluations were performed independently for each of the manufacturing tolerances and the associated reactivity uncertainties were combined statistically. All calculations were made for an explicit modeling of the fuel and Report: HI-2043321 I Project 1430 Proprietary Information Denoted by Shaded Text

storage cell to define the limiting enrichment-burnup combinations for spent fuel that assures the safe storage of spent fuel in the BWR racks.

The criticality safety criteria used in the analysis was (1) the racks remain Subcritical without any credit for the soluble boron present and (2) partial credit is taken for the soluble boron to assure the reactivity remains below 0.95 under normal and accident conditions.

The maximum klff values were determined assuming an infinite radial array of storage cells with a finite axial length, water reflected. For each initial enrichment, a minimum burnup was determined that assures the maximum keff, including calculational and manufacturing uncertainties, remains sub-critical under the assumed absence of all soluble boron.

A conservative axial burnup distribution (Figure 1) was used in the calculations. Figures 2 and 2a summarize the results of these analyses, showing the minimum acceptable burnup for fuel of various initial maximum planar average enrichments at cooling times of 4 and 7 years, respectively. The limiting points in Figures 2 and 2a may be fitted by the following polynomial functions of the initial maximum planar average enrichment, E.

4 years: Burnup Limit = 0.5406

  • E3 - 5.3804
  • B2 + 33.389
  • E - 38.167 7 years: Burnup Limit = 0.3850
  • E3 - 3.8506 E2 + 28.063
  • E - 32.905 The polynomials were selected so all points are bounded. Table 5 shows the calculated minimum burnups and the values determined using the polynomials for both cooling times.

The soluble boron concentration required to maintain keff below 0.95, including all manufacturing and calculation tolerances, for storage of fuel in the pool was determined to be 300 ppm under normal conditions. An accident scenario, where a maximum reactivity fuel assembly is accidentally loaded into an otherwise filled rack was also evaluated. For this accident case, 325 ppm soluble boron would be required to maintain kff below 0.95. This minimum soluble boron concentration of 325 ppm is well below the 2000 ppm soluble boron administrative limit for the pool water. This provides a large safety margin in reactivity.

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Based on the analyses presented herein, it is concluded that the BWR Boraflex spent fuel storage racks can safely accommodate fuel with initial maximum planar average enrichments up to 4.6%, with assurance that the maximum reactivity, including calculational and manufacturing uncertainties, will be less than 0.95, with 95% probability at the 95% confidence level, provided only that (1) the fuel conforms to the enrichment-burnup limits for the spent fuel as depicted in Figures 2 and 2a, and (2) that a minimum of 325 ppm soluble boron is maintained. The limiting burnups shown in Figures 2 and 2a, and in Table 5 are the assembly average burnups and, in their application, must be adjusted for the plant's uncertainty in determining the actual burnups of the spent fuel assemblies.

2.0 ANALYSIS CRITERIA AND ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative analysis criteria or assumptions were used.

  • Criticality safety analyses were based upon an infinite radial array of cells; i.e. no credit was taken for radial neutron leakage.
  • Minor structural materials were neglected; i.e. spacer grids were conservatively assumed to be replaced by water.
  • Because the temperature coefficient of reactivity is positive in the absence of Boraflex, the analyses assumed the administrative limit temperature of 150 TF. Higher temperatures would be an accident condition for which soluble boron credit is permitted.

" The axial burnup distribution calculations were performed assuming an axial distribution shown in Figure 1 (and Table 6).

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3.0 ACCEPTANCE CRITERIA The primary acceptance criterion under normal conditions is that the maximum keff shall be less than critical, including calculational uncertainties and effects of mechanical tolerances under the postulated absence of all soluble boron. Partial credit is taken for the soluble boron in the pool water to assure that the maximum keff shall be less than 0.95, including calculational uncertainties and effects of mechanical tolerances, under normal and accident conditions.

Applicable codes, standards, and regulations, or pertinent sections thereof, include the following:

" General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

  • Code of Federal Regulation 10CFR50.68, Criticality Accident Requirements.
  • USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications.

December 1981.

  • ANSI-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
  • L. Kopp, "Guidance On The Regulatory Requirements For Criticality Analysis of.Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum L. Kopp to Timothy Collins, August 19, 1998.

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4.0 DESIGN AND INPUT DATA 4.1 FUEL ASSEMBLY DESIGN The design basis fuel assembly is the GE 13 assembly, a 9x9 array of U0 2 fuel rods and 2 large water rods. Table 2 provides the pertinent design details for the GE13 fuel assembly as well as other fuel designs stored in the Harris spent fuel pools. Tables 7 and 7a provide comparisons of the reactivities of the fuel assembly designs present in the spent fuel pool for enrichments between 2.0 wt% and 4.6 wt% 235U for burnups above and below the minimum burnup requirement. Table 7 shows that for a cooling time of 4 years the GE 3, GE4 and GE 7 assemblies have higher reactivities than the GEl3 assembly. However, these fuel assembly designs were used in earlier operation of the Brunswick reactor and these assemblies have cooling times in excess of 26, 23 and 12 years for the GE3, GE4 and GE7 assemblies, respectively. Table 7 shows that when cooling time is also considered for these fuel assembly designs, the GE13 assembly with 4 years cooling time has the highest reactivity. Table 7a shows a similar comparison for 7 years cooling time for the GE13 assembly. Therefore, the GE13 assembly is used in all further calculations described in subsequent sections.

The axial dimensions of the GEl 3 fuel assembly are:

GE-13 Bundle Axial Description Height Lattice Description Number of fuel rods (from bottom of bundle, in inches) 0.0 to 6.0 Natural Uranium Blanket 74 fuel rods 6.0 to 108.0 Full Lattice 74 fuel rods 108.0 to 138.0 Part-Length Lattice 66 fuel rods 138.0 to 146.0 Natural Uranium Blanket 66 fuel rods 4.2 CORE OPERATION AND CONTROL RODS Core operating parameters are necessary for fuel depletion calculations performed with CASMO-4. The core parameters used for the depletion calculations are presented in Table 2a.

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Temperature and void fraction values are taken as the upper bound (most conservative) of the core operating parameters of Brunswick. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity values.

Additionally, the Brunswick reactor uses control rod blades for reactor and power control during operation. The geometrical and material properties of the control rods are provided in Table 2b.

The control rod operating strategy at Brunswick for GE 13 fuel did not allow fresh fuel to be placed in a core location that would have planned control rod insertion. Therefore, control rod insertion is limited to once and twice burned fuel assemblies. Typical fuel is controlled for 3 GWD/MTU intervals then uncontrolled for 3 GWD/MTU intervals. Therefore, to conservatively bound any control rod insertion, fuel assemblies are modeled with an initial interval of 12 GWD/MTU (i.e., first cycle) of uncontrolled operation followed by intervals of 3 GWD/MTU controlled and uncontrolled operation. This is conservative as fuel is not actually controlled for 3 GWD/MTU as the flux suppression of the control rod blade significantly decreases exposure accumulation.

4.3 STORAGE RACK DESIGN The storage cells are composed of stainless steel boxes, joined at the corners in an egg-crate structure. Initially the design included Boraflex as the absorber, although in the present analyses, the Boraflex is assumed to be lost. The storage cells are located on a lattice spacing of 6.25 inches (for conservatism, a lattice spacing of 6.22 inches was used in the analyses). The box wall thickness is 0.075 inches, and the box inside dimension is 6.05 inches. The wrapper wall thickness is 0.035 inches. The Boraflex panels were assumed to be completely replaced by water. A cross-section of the storage cell is shown in Figure 3.

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5.0 METHODOLOGY 5.1 GENERAL DESCRIPTION The primary criticality analyses (at 95% probability, 95% confidence level)(') were performed with the three-dimensional MCNP4a code(2). Benchmark calculations (Appendix A) have determined a calculational bias of 0.0012+0.0090 Ak for MCNP4a calculations.

CASMO4, a two-dimensional deterministic code(3) using transmission probabilities, was used to evaluate the small (differential) reactivity effects of manufacturing tolerances and to determine nuclide concentrations developed in the depletion calculations.

In the geometric model used in the calculations, each fuel rod and each fuel assembly were explicitly described. Reflecting boundary conditions effectively defined an infinite radial array of storage cells. In the axial direction, a 30-cm water reflector was used to conservatively describe axial neutron leakage. Each stainless steel box and water gap were also described in the calculational model. The fuel cladding material was zirconium.

MCNP4a Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To assure convergence and to minimize the statistical uncertainty of the calculated reactivities, a minimum of 4 million neutron histories was accumulated in each calculation, generally resulting in a statistical uncertainty of about -0.0003 Ak.

5.2 AXIAL BURNUP DISTRIBUTION Usually BWR storage rack analyses do not credit soluble boron in the pool water and use low burnup fuel with credit for gadolinia. However, in the Harris spent fuel pool, soluble boron is present and higher burnup fuel is necessarily credited. With high burnup fuel, the axial distribution in burnup becomes important and must be considered in any assessment of reactivity.

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Progress Energy provided 16 axial burnup distributions for assemblies of approximately 4.03 wt% enrichment with burnups between 30.2 GWD/MTU and 45.9 GWD/MTU. The minimum required burnup for assemblies with 4.0 wt% enrichment as shown in Table 5 is 44.0 GWD/MTU. Therefore, four axial burnup distributions with an assembly average burnup near or below the minimum required burnup specified in Table 5 were chosen and averaged to determine an appropriate axial burnup distribution.

The axial burnup distribution is conservative because the profiles selected for averaging are from assemblies with burnups near or below the required minimum burnup. Assemblies with much lower burnups, which would have profiles that might produce higher reactivities, would not be able to be stored in the racks because they would not meet the burnup requirement. Assemblies with higher burnups have profiles that are more cosine shaped, which would further reduce the reactivity. This is shown in the following table where the resultant reactivity of an axial burnup distribution from an assembly with an enrichment of approximately 4.0 wt% with a burnup of 45.9 GWD/MTU is compared to the resultant reactivity from an assembly with the axial burnup distribution in Table 6. These calculations were performed at a burnup of 45 GWD/MTU with an enrichment of 4.0 wt% 235U.

Profile Table 6 YJM160 Assembly Burnup 37.51 45.88

[GWD/MTU]

Calculated keff 0.9554 0.9186 Therefore, based on the level of conservatism inherent in choosing the axial burnup distribution in the manner described above, it is not necessary to confirm that the axial burnup distributions of individual assemblies are bounded by the assumed axial burnup distribution. Additionally, other areas of conservatism such as the assumed reactor records burnup uncertainty, longer cooling times than the assumed 4 or 7 years, and the presence of soluble boron further insure that the BWR racks meet the acceptance criteria stated in Section 3.0.

The assembly burnup for the burnup profile in Table 5 is an average of the assembly average burnups from the four assemblies selected to determine the axial burnup profile in Table 5.

2 This is the calculated keff from Table I for 4.0 wt% at 45 GWD/MTU Report: HI-2043321 8 Project 1430 Proprietary Information Denoted by Shaded Text

In the present analyses, the final axial burnup distribution assumed is illustrated in Figure 1, as determined by 25 segment calculations. Four different axial burnup distributions for GEl 3 fuel were selected. The normalized axial burnup distributions are very nearly the same. An average of the four axial burnup distributions for GEl3 fuel was developed and used in the present analyses as illustrated in Figure 1 and listed in Table 6.

These calculations use axial blankets of natural uranium oxide and include the effect of part-length fuel rods. The lower fully rodded zone (up to 108 inches above the bottom of the fuel) contains 74 fuel rods and the partially-rodded zone contains 66 fuel rods. At the top of the assembly, the blanket is 8 inches in length and the bottom blanket is 6 inches long. The top axial blanket is extended further 4 inches (to make the total fuel height 150 inches) so that the axial burnup distribution and the fuel height modeled are consistent.

Separate CASMO4 depletion calculations were made for each of the 25 axial segments and the isotopic composition of each of the 25 segments were transferred to a 3-dimensional MCNP4a case, thereby inherently incorporating the effect of the axial burnup distribution.

In some cases of lower average burnups, a uniform axial burnup distribution can result in a higher reactivity than the distributed burnup case. Therefore, all calculations were also performed with an axially flat profile equal to the assembly average burnup. The profile (distributed or flat) which produced the highest reactivity was used to determine the maximum keff in the storage racks.

6.0 ANALYSIS RESULTS 6.1 EVALUATION OF TOLERANCE UNCERTAINTIES CASMO4 calculations were made to determine the uncertainties in reactivity associated with manufacturing tolerances. Results of these calculations are shown in Tables 1, la and 3. The reactivity effects were separately evaluated, in a sensitivity study for each independent tolerance, Report: HI-2043321 9 Project 1430 Proprietary Information Denoted by Shaded Text

and the results combined statistically, using the root mean square methodology. Tolerances considered include the following:

, Tolerance in pitch - minimum pitch (6.22") used which is less than the nominal pitch (6.25")

  • Tolerance in Steel Box I.D.
  • Tolerance in Stainless Steel Box Wall Thickness
  • Tolerance in U0 2 Density In addition to these mechanical tolerances, uncertainties due to tolerance in benchmarking calculations, statistical variation in MCNP4a calculations and the estimated uncertainty/in depletion calculations (section 6.1.1 below) are included (See Tables 1 and 1a). Also included is the incremental reactivity between 20 TC (MCNP4a calculation) and the reference temperature of 150 TF. The effect of soluble boron on the manufacturing tolerances is not considered in the BWR racks since the soluble boron requirement is low (less than 325ppm) compared with the plant Technical Specification.

6.1.1 UNCERTAINTY IN DEPLETION CALCULATIONS The uncertainty in depletion calculations were taken as 5% of the reactivity decrement from beginning-of-life to the burnup of concern(4). These uncertainties depend on burnup and are listed in Table 1 and la.

6.2 ABNORMAL AND ACCIDENT CONDITIONS A brief summary of the calculated reactivity effects of the accident conditions is given in Table 4.

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6.2.1 ECCENTRIC LOCATION OF FUEL ASSEMBLIES The fuel assemblies are normally stored in the center of the storage cells. Calculations were made with the fuel assemblies assumed to be in the corner of the storage rack cell (eccentric positioning of a four-assembly cluster at closest approach). As shown in Figure 3, reflective boundary conditions are used on the periphery of the single cell model. This creates an infinite array of storage cells. By moving the assembly in the single storage cell model to the corner of the storage cell, an infinite array of storage cells is created with each cluster of four assemblies being placed closest to a common corner. Eccentric positioning of spent fuel assemblies resulted in a slightly lower reactivity in these racks.

6.2.2 TEMPERATURE AND VOID EFFECTS Temperature effects were also evaluated using CASMO4. These results presented in Table 8 show that the temperature coefficient of reactivity is positive and the reactivity at 150 'F (maximum expected spent fuel pool water temperature) is used to derive a bias. Any residual Boraflex that might remain would reduce the temperature penalty. At the submerged depth of the storage pool, the maximum temperature at boiling is 120 'C, and the void coefficient of reactivity is negative.

The reference MCNP4a calculations were performed at a water density corresponding to a temperature of 20 TC. The reactivity increment between that at the maximum expected water temperature and at 20 °C is taken into account as an additive term (bias). This bias, at each enrichment considered, is listed in Table I and la.

6.2.3 MIS-LOADED FUEL ASSEMBLY ACCIDENT The potential effects of a fuel mis-loading accident condition were also considered in this study.

To evaluate the consequence of the fuel mis-loading accident, the misloaded fuel assembly was Report: HI-2043321 11 Project 1430 Proprietary Information Denoted by Shaded Text

assumed to be the most reactive assembly possible - a spent fuel assembly with a k., of 1.333 in the Standard Cold Core Geometry.* Fuel isotopes (not including any gadolinia) were extracted from the CASMO4 calculation and used in a 3-dimensional MCNP4a calculational model. All other assemblies in the calculation used the CASMO4 nuclide inventory for the normal fuel burnup.

For the most serious postulated mis-loading accident scenario, calculations were performed to determine the soluble boron concentration required to maintain keff below 0.95 in the pool under the postulated accident scenario. Results of this calculation showed that 325 ppm soluble boron would be adequate to assure a keff less than 0.95.

7.0 CONCLUSION

S Fuel assemblies with spent fuel having at least the burnup-enrichment combination as depicted in Figures 2 and 3, and Table 5 may be safely accommodated in the storage racks, with no other constraints on their placement in the pool.

3 The SCCG kinf of 1.33 is based on the maximum SCCG kinf provided in Table 2 (1.32) plus an additional 0.01 Ak adder to account for differences in the calculation of the SCCG kinf between Holtec and the fuel vendor.

. The k. (SCCG) is defined as the reactivity of an infinite array of assemblies at 20 'C on a 6-inch lattice pitch without void or any control element.

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8.0 REFERENCES

[1] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

[2] J. F. Briesmeister, editor, MCNP - A General Monte Carlo N-Particle Transport Code, LA-12625-M, Los Alamos National Laboratory, November 1993.

[3] A. Ahlin, M. Edenius, H. Haggblom, "CASMO- A Fuel Assembly Burnup Program,"

AE-RF-76-4158, Studsvik report (proprietary).

A. Ahlin and M. Edenius, "CASMO- A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604, 1977.

D. Knott, "CASMO4 Benchmark Against Critical Experiments", Studsvik Report SOA-94/13 (Proprietary).

M. Edenius et al., "CASMO4, A Fuel Burnup Program, Users Manual" Studsvik Report SOA/95/1.

[4] L. Kopp, "Guidance On The Regulatory Requirements For Criticality Analysis Of Fuel Storage At Light-Water Reactor Power Plants", USNRC Internal Memorandum, L. Kopp to Timothy Collins, August 19, 1998.

[5] "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit,"

NUREG/CR-6760, ORNL/TM-2000/23 1, March 2002.

[6] HI-2094370R0, "CASMO-4 Benchmark for Spent Fuel Pool Criticality Analysis."

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Table 1 : Maximum Reactivity at Various Initial Enrichment - Minimum Burnup Combinations (4 years cooling)

Initial Enrichment, wt% U-235 1.50% 2.00% 2.50% 3.00% 3.50% 4.00% 4.60%

Minimum Burnup,MWD/KgU 1.5 11.1 20.1 27.6 35.5 43.9 53.8 MCNP Bias 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 Temperature Correction to 0.0044 0.0061 0.0067 0.0069 0.0071 0.0071 0.0071 150°F Calculated kff 0.9604 0.9636 0.9634 0.9630 0.9615 0.9609 0.9600 Uncertainties MCNP Bias Uncertainty 0.0090 0.0090 0.0090 0.0090 0.0090 0.0090 0.0090 CASMO Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculational 0.0006 0.0006 0.0006 0.0006 0.0006 0.0008 0.0006 Eccentricity 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 Manufact. Tolerances 0.0219 0.0157 0.0129 0.0118 0.0108 0.0099 0.0094 Depletion Uncertainty 0.0006 0.0054 0.0088 0.0115 0.0137 0.0154 0.0171 Total Uncertainty 0.0238 0.0191 0.0183 0.0190 0.0198 0.0206 0.0216 Maximum keff 0.9898 0.9900 0.9895 0.9900 0.9896 0.9897 0.9900 Report: HI-2043321 . 14 Project 1430 Proprietary Information Denoted by Shaded Text

Table I a : Maximum Reactivity at Various Initial Enrichment - Minimum Burnup Combinations (7 years cooling)

Initial Enrichment, wt% U-235 1.50% 2.00% 2.50% 3.00% 3.50% 4.00% 4.60%

Minimum Burnup,MWD/KgU 1.4 10.5 19.2 26.2 34.2 42.3 51.7 MCNP Bias 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 Temperature Correction to 0.0044 0.0060 0.0071 0.0067 0.0074 0.0070 0.0070 150°F Calculated klff 0.9605 0.9633 0.9633 0.9629 0.9615 0.9608 0.9601 Uncertainties MCNP Bias Uncertainty 0.0090 0.0090 0.0090 0.0090 0.0090 0.0090 0.0090 CASMO Bias Uncertainty 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 0.0025 Calculational 0.0006 0.0006 0.0006 0.0006 0.0006 0.0006 0.0006 Eccentricity 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 Manufact. Tolerances 0.0219 0.0159 0.0126 0.0118 0.0106 0.0100 0.0094 Depletion Uncertainty 0.0006 0.0054 0.0089 0.0115 0.0137 0.0154 0.0171 Total Uncertainty 0.0239 0.0192 0.0180 0.0190 0.0197 0.0206 0.0216 Maximum keff 0.9899 0.9897 0.9896 0.9898 0.9898 0.9897 0.9899 Report: HI-2043321 15 Project 1430 Proprietary Information Denoted by Shaded Text

Table 2: BWR Fuel Characteristics FuelGE 3 GE 4 GE 7 GE 8 GE 9 GE 10 GE 13 Assembly _ G NOTE: All dimensions in inches Clad O.D. 0.563 0.493 0.483 0.483 0.483 0.483 0.440 Clad I.D. 0.489 0.425 0.419 0.419 0.419 0.419 0.384 Clad Material Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Pellet Diamet 0.477 0.416 0.410 0.411 0.411 0.411 0.376 Diameter Stack Density 10.31 10.40 10.54 10.58 10.54 10.54 10.54 Maximum Enrichm 4.6 4.6 4.6 4.6 4.6 4.6 4.6 Enrichment SCCG kinf *1.32 *_1.32 *<1.32 *1.32 *1.32 *1.32 *1.32 Active Acti Fuel 144 146 150 150 150 150 146 Length_____

Axial Length of Partial 108 Rods, in Fuel Rod 7x7 8x8 8x8 8x8 8x8 8x8 9x9 Array Number Fuel 49 63 62 60 60 60 74/66 Rods Fuel Rod 0.738 0.640 0.640 0.640 0.640 0.640 0.566 Pitch Number of WaterWtros Rods 0 1 2 4 1 1 2 Water Rod NA 0.493 0.591 0.591/ 1.34 1.34 0.980 O.D. 0.483 WaterRod NA 0.425 0.531 0.531/ 1.26 1.26 0.920 I.D. 0.431 Channel I.D. 5.278 5.278 5.278 5.278 5.278 5.278 5.278 Channel Thicnes 0.080 0.080 0.080 0.080 0.080 0.070 0.070 Thickness Notes:

1. The GE 13 assembly has 8 part length rods, 108 inches in height.
2. The GE 5 and GE 6 are identical to the GE 7 for the fuel parameters listed.
3. This data was provided by Progress Energy. The enrichment is the maximum planar average enrichment.

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Table 2a: Brunswick Core Operating Parameters for Depletion Parameter Value Reactor Specific Power, MW/MTU 30 Maximum Core Plane Average Fuel 1260 Temperature, 'F Core Average Moderator Temperature at the 560 Top of the Active Region, 'F In-Core Assembly Pitch, Inches 6.00 Maximum Void Fraction 0.77 Report: HI-2043321 17 Project 1430 Proprietary Information Denoted by Shaded Text

Table 2b: BWR Control Rod Characteristics Parameter Value Central Support Tie Rod Span, inches 1.56 Wing Tip to Wing Tip Blade Span, inches 9.81 Blade Wing Thickness, inches 0.312 Blade Sheath Thickness, inches 0.056 Poison Tube Wall Thickness, inches 0.025 Poison Tube OD, inches 0.188 Absorber Zone Axial Length, inches 143 Number of B 4C Tubes (Per Wing) 21 B 4C Density, g/cc 1.76 Report: HI-2043321 18 Project 1430 Proprietary Information Denoted by Shaded Text

Table 3: Tolerances and Their Reactivity Effects Description Tolerance Reactivity Effect, Ak

1) Tolerance in StorageSeTalI Box IDSeTal1
2) Tolerance in StainlessSe Tal I Steel Wall Thickness..SeTal1
3) Tolerance Enrichment in Uranium g igTbl
4) T olerance in U 020 . 0Se Ta l I Density
5) Uncertainty due to FuelNAegtv EccentricityNAegtv
6) Uncertainty in Depletion 5% of Reactivity Decrement See Table 1 Calculations I I Report: HI-2043321 19 Project 1430 Proprietary Information Denoted by Shaded Text

Table 4: Summary of Abnormal/ Accident Conditions Condition Consequence 150 TF used for normal storage condition.

Temperature Increase Higher temperatures are accident conditions with credit for soluble boron allowed.

Void (Boiling) Negative void coefficient of reactivity Assembly Drop on Top of Rack Negligible Seismic Movement Negligible Mis-Loaded Fuel Assembly Requires 325 ppm soluble boron Report: HI-2043321 20 Project 1430 Proprietary Information Denoted by Shaded Text

Table 5: BWR Boraflex Racks Burnup Versus Enrichment Requirement Initial Maximum 4 years cooling 7 years cooling Planar Average Calculated Calculated Enrichment, wt% Burnup Limit m ial Burnup Limit Polynomial Fit, U-235 MWD/KgU MWD/KgU MWD/KgU MWD/KgU 1.5 1.5 1.6 1.4 1.8 2.0 11.1 11.4 10.5 10.9 2.5 20.1 20.1 19.2 19.2 3.0 27.6 28.2 26.2 27.0 3.5 35.5 36.0 34.2 34.7 4.0 43.9 43.9 42.3 42.4 4.6 53.8 54.2 51.7 52.2 Report: HI-2043321 21 Project 1430 Proprietary Information Denoted by Shaded Text

Table 6: Average Segment Axial Burnup Distribution Segment Axial Height from Top Relative Burnup Number of Active Fuel (cm) 1 0-15.24 0.1740 2 15.24 - 30.48 0.2490 3 30.48 -45.72 0.5560 4 45.72 - 60.96 0.7160 5 60.96 - 76.20 0.8600 6 76.20 - 91.44 0.9610 7 91.44 - 106.68 1.0290 8 106.68 - 121.92 1.0370 9 121.92 - 137.16 1.0860 10 137.161- 152.40 1.1260 11 152.40 - 167.64 1.1570 12 167.64 - 182.88 1.1820 13 182.88 - 198.12 1.2050 14 198.12 - 213.36 1.2260 15 213.36 - 228.60 1.2430 16 228.60 - 243.84 1.2590 17 243.84 - 259.08 1.2720 18 259.08 - 274.32 1.2840 19 274.32 - 289.56 1.2930 20 289.56 - 304.80 1.2970 21 304.80 - 320.04 1.2930 22 320.04 - 335.28 1.2490 23 335.28 - 350.52 1.1300 24 350.52 - 365.76 0.8680 25 365.76 - 381.00 0.2530 Report: HI-2043321 22 Project 1430 1 Proprietary Information Denoted by Shaded Text

Table 7: Assembly Design Reactivities (4 years)

Burnup GE3 GE3 GE4 GE4 GE7 GE7 GE8 s GE9 GEIO GE13

[GWD/MTU] 4 years 26 years 4 years 23 years 4 years 12 years 4 years 4 years 4 years 4 years Enrichment = 2.0 wt% 235U 10 0.9891 0.9719 0.9832 0.9674 t 0.9811 0.9726 0.9787 0.9777 0.9642 0.9794 12.5 0.9660 0.9418 0.9601 0.9378 0.9574 0.9454 0.9546 0.9536 0.9398 0.9555 Enrichment = 3.0 wt% 235 U 25 0.9714 0.9318 0.9656 0.9291 0.9642 0.9442 0.9629 0.9618 0.9462 0.9628 27.5 0.9515 0.9071 0.9456 0.9047 0.9435 0.9212 0.9418 'J 0.9407 0.9249 0.9421 2 35 Enrichment = 4.0 wt% U 40 0.9445 0.8925 0.9388 0.8908 t 0.9377 0.9113 0.9372 0.9361 0.9193 0.9369 42.5 0.9276 0.8722 0.9219 0.8709 0.9202 0.8919 0.9191 0.9180 0.9010 0.9192 Enrichment = 4.6 wt% 23SU 50 0.9220 0.8639 0.9164 0.8628 0.9151 0.8853 0.9146 0.9135 0.9070 0.9145 52.5 0.9068 0.8460 0.9013 0.8451 0.8993 0.8681 0.8983 0.8971 0.8903 0.8985 Report: HI-2043321 23 Project 1430 Proprietary Information Denoted by Shaded Text

Table 7a: Assembly Design Reactivities (7 years)

Burnup GE3 GE4 GE7 GE8 GE9 GEl 0 GEl 3

[GWD/MTU] 27 years 24 years 13 years 11 years 11 years 9 years 7 years Enrichment = 2.0 wt% 235U 10.0 0.9715 0.9669 0.9717 0.9711 0.9702 0.9587 0.9759 12.5 0.9413 0.9372 0.9442 0.9440 0.9429 0.9319 0.9505 Enrichment = 3.0 wt% 23SU 25.0 0.9309 0.9280 0.9423 0.9451 I0.9441 0.9331 0.9544 27.5 0.9061 0.9035 0.9190 0.9218 0.9207 0.9100 0.9326 Enrichment = 4.0 wt% 23SU 40.0 0.8913 0.8894 0.9087 0.9134 0.9124 0.9017 0.9257 42.5 0.8710 0.8694 0.8892 0.8937 0.8926 0.8821 0.9072 Enrichment = 4.6 wt% 235U 50.0 0.8626 0.8612 0.8825 0.8879 0.8868 0.8870 0.9018 52.5 0.8446 0.8435 0.8651 0.8702 0.8690 0.8692 0.8852 Report: HI-2043321 24 Project 1430 Proprietary Information Denoted by Shaded Text

Table 8: Reactivity Effect of Temperature and Void Content 4.0 wt%, 45.0 MWD/kgU Temperature [°C] Ak 0 Reference 4 + 0.0007 26.9 + 0.0049 65.6 (150 °F) +0.0115 123 + 0.0214 123, 10% Void + 0.0176 Report: HI-2043321 25 Project 1430 Proprietary Information Denoted by Shaded Text

Axial Burnup Distribution 1.4 1.2 E 0.8 . ..

0.

0.4 02 0 5 10 15 20 25 30 Node (from Top)

Figure 1: Axial Burnup Distribution Report: HI-2043321 26 Project 1430 Proprietary Information Denoted by Shaded Text

Fig. 2: Minimum Fuel Burnup for Acceptable Storage of Spent Fuel of Various Initial Planar Average Enrichments (4 years).

Report: HI-2043321 27 Project 1430 Proprietary Information Denoted by Shaded Text

Boraflex Racks Loading Curve 60

.50

"* 40 +V 0 30 20 u 20 E

.~10 0

1.5 2 2.5 3 3.5 4 4.5 5 Initial Maximum Planar Average Enrichment (wt%U-235)

Fig. 2a: Minimum Fuel Burnup for Acceptable Storage of Spent Fuel of Various Initial Planar Average Enrichments (7 years).

Report: HI-2043321 28 Project 1430 Proprietary Information Denoted by Shaded Text

6.22' SQ. LATTICE SPACING REFLECTIVE BOUNDARY CONDITION z z 0 z C-,

z 6.05' SQ, z

Z)

Lu Li

-J -J

.... 77 ....

REFLECTIVE BOUNDARY CONDITION DETAIL 'A' BOX WALL THICKNESS + SHEATHING THICKNESS )/2

=( 0.075 + 0,035 )12 = 0.055' W77~Z77Z7Z7~

t REFLECTIVE BOUNDARY CONDITION BORAFLEX GAP THICKNESS )/2

= ( 0.060 )/2 = 0.030 DETAIL 'A' Fig. 3: Cross-Section of Typical Storage Cell (Calculational Model)

Report: HI-2043321 29 Project 1430 Proprietary Information Denoted by Shaded Text

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