ML100130545
| ML100130545 | |
| Person / Time | |
|---|---|
| Site: | Limerick, 06300001  |
| Issue date: | 08/11/2003 |
| From: | Radulescu H US Dept of Energy (DOE) |
| To: | NRC/NMSS/DHLWRS |
| Shared Package | |
| ML100110027 | List: |
| References | |
| DOC.20030825.001 CAL-DSU-NU-000002 Rev 00A | |
| Download: ML100130545 (42) | |
Text
{{#Wiki_filter:DOC.20030825.0001
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- 2. Page I
- 3. System . ".. *'1::*4:;'~()ocUnienflde*l'ltifje.r Uncanistered Spent Nuclear F u e l * .'.: :::GA:L~Q.sU-NU-{)00002 REV OOA
- 5. Title . ~?;.'
Limerick Unit 1 Radiochemical Assay Comp~sons to SA*S2H Calculations
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- 7. Document Status Designation
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- 8. Notes/Comments Attachments Total Number ~f PageS 3
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Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 2 of 37 CONTENTS Page
- 1. PURPOSE ..................................................................................................................................5
- 2. METHOD ..................................................................................................................................6
- 3. ASSUMPTIONS........................................................................................................................7
- 4. USE OF COMPUTER SOFTWARE AND MODELS..............................................................8 4.1 SOFTWARE APPROVED FOR QA WORK....................................................................8 4.1.1 SAS2H .....................................................................................................................8 4.1.2 MCNP ......................................................................................................................8
- 5. CALCULATION .......................................................................................................................9 5.1 CALCULATION METHOD..............................................................................................9 5.1.1 SAS2H Fuel Depletion Description.........................................................................9 5.2 SAMPLE DESCRIPTION ...............................................................................................10 5.3 FUEL ASSEMBLY DESIGN AND OPERATING PARAMETERS .............................13 5.3.1 SAS2H Material Specifications .............................................................................14 5.3.2 SAS2H Operating History Specifications..............................................................16 5.4 SAS2H PATH B REPRESENTATIONS.........................................................................23 5.5 MCNP SPECIFICATIONS ..............................................................................................25
- 6. RESULTS ................................................................................................................................30 6.1 SAMPLE RESULTS FROM ASSEMBLY YJ1433........................................................31 6.2 MCNP CALCULATION RESULTS ...............................................................................32
- 7. REFERENCES ........................................................................................................................34 7.1 DOCUMENTS CITED ....................................................................................................34 7.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES ...............................35 7.3 SOURCE DATA LISTED BY DATA TRACKING NUMBER .....................................36 7.4 SOFTWARE CODES ......................................................................................................36
- 8. ATTACHMENTS....................................................................................................................37
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 3 of 37 FIGURES Page 5-1. SAS2H General Path B Model used for Assembly YJ1433................................................. 24
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 4 of 37 TABLES Page 5-1. Fuel Sample Characteristic Parameters ................................................................................ 11 5-2. Measured Concentrations for Assembly YJ1433 Samples .................................................. 12 5-3. YJ1433 Fuel Assembly Geometric and Material Information ............................................. 13 5-4. Nominal Characteristics of Sampled Fuel Rods................................................................... 13 5-5. Trace Isotopes Specified in Fresh Fuel Compositions ......................................................... 14 5-6. Zircaloy-2 and Zircaloy-4 Compositions ............................................................................. 14 5-7. SAS2H Compositions Specific to Assembly YJ1433 .......................................................... 16 5-8. Constants for Fuel Temperature Calculations ...................................................................... 17 5-9. Operating History Information for Assembly YJ1433 ......................................................... 21 5-10. Material Specifications for SB-575 N06022 ...................................................................... 27 5-11. Material Specifications for SS316NG ................................................................................ 27 5-12. Material Specifications for Neutronit A978 with 1.62 wt% Boron.................................... 28 5-13. Material Specifications for Al 6061 ................................................................................... 28 5-14. Material Specifications for Grade 70 A516 Carbon Steel.................................................. 28 5-15. Material Specifications for Zircaloy-2 ............................................................................... 28 5-16. Material Specifications for Zircaloy-4 ............................................................................... 29 6-1. SAS2H Calculated Isotopic Concentrations for Assembly YJ1433 Samples ...................... 31 6-2. Assembly YJ1433 Sample Percent Differences ................................................................... 32 6-3. MCNP Results for Fresh Nuclear Fuel................................................................................. 33 6-4. MCNP Results for Spent Nuclear Fuel Samples .................................................................. 33
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 5 of 37
- 1. PURPOSE The objective of the Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations is to determine the accuracy of the SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN: 10129-4.4A-00), in predicting the isotopic concentrations of spent fuel, and to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on the system reactivity. The scope of this calculation covers eight different spent fuel samples from a fuel assembly that was irradiated in the Limerick Unit 1 boiling water reactor (BWR). The spent fuel samples evaluated are from a three-cycle burn period, representing burnups from 37.02 GWd/MTU through 65.54 GWd/MTU (Reager 2003).
This report is an engineering calculation supporting the development of validation reports to be used for License Application of the proposed Monitored Geologic Repository (MGR), and was performed under Administrative Procedure-3.12Q, Design Calculations and Analyses. This calculation is subject to the Quality Assurance Requirements and Description (DOE 2003) per the activity evaluation under work package number ACRM01 in the technical work plan TWP-EBS-MD-000014 REV 00 (BSC 2002). The control of the electronic management of data was accomplished in accordance with methods specified in BSC (2002).
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 6 of 37
- 2. METHOD The calculational method used to determine the spent nuclear fuel isotopics consisted of using the SAS2H control sequence of the baselined modular code system SCALE to deplete the fuel for two initial fuel enrichments and eight burnups. The isotopic predictions are then compared against measured concentrations from the fuel assembly that was represented in the depletion calculations to determine the accuracy of the predicted values.
The analytical methods employed for this evaluation were the SAS2H control module of the baselined modular code system SCALE and baselined code system MCNP, Version 4B2LV (CSCI: 30033-V4B2LV). Based upon fuel assembly design, power history, and operating data for the specific assembly in the Limerick Unit 1 core, a computational representation was developed for use with SAS2H. The SAS2H module is used to perform a fuel depletion analysis to predict the isotopic concentrations in localized areas of assembly pins. The isotopic concentrations predicted by the SAS2H module are then compared with measured concentrations of the same localized areas (axial locations) of the assembly pins to determine the accuracy of the developed calculational representation. The measured and calculated isotopic compositions from SCALE were then used as input to the baselined MCNP code to calculate the neutron multiplication factor in order to quantify the overall effect that the variations between measured and calculated isotopic concentrations have on system neutron multiplication. The measured isotopic concentrations used for comparisons in this evaluation are taken from Reager (2003).
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 7 of 37
- 3. ASSUMPTIONS 3.1 It is assumed that the YJ1433 fuel assembly channel was made of Zircaloy-4. Although specific information about the YJ1433 assembly channel material could not be found, various sources indicate that the material used for the assembly channel in other similar General Electric fuel designs was Zircaloy-4 (CRWMS M&O 2000a, Table S2.6.4 and BSC 2001, p. 25). This assumption is used in Section 5.
3.2 It was assumed that the 234U wt% in the fresh fuel composition was approximately 0.9 % of 235 U wt%. The rationale for this assumption is that no information about the 234U content in the fresh fuel rods in YJ1433 assembly at the Limerick Unit 1 BWR is available, therefore the information about 234U and 235U contents in similar fuel types is used to approximate the correlation between the wt% contents for the two isotopes (CRWMS M&O 2000a, p. S2.6.11 and BSC 2001, p. 15). This assumption is used in Section 5. 3.3 It is assumed that there is no 236U in the fresh fuel composition. The rationale for this assumption is that no information about the 236U content in the fresh fuel rods in YJ1433 assembly at the Limerick Unit 1 BWR was available, and 236U is only present in 235U that has been obtained from reprocessed uranium. This assumption is used in Section 5. 3.4 The heat transfer between the outer surface of the fuel rod and the coolant is assumed to be completely efficient (i.e., there is no temperature jump across the boundary layer). The rationale for this assumption is that is reasonable for boiling heat transfer and should have an acceptably small effect on fuel temperature. This assumption is used in Section 5. 3.5 The fuel thermal conductivity is assumed to be invariant with radial temperature distribution and fuel exposure. The rationale for this assumption is that since the fuel temperature is lumped in the SAS2H depletion model, this is acceptable. This assumption is used in Section 5. 3.6 It is assumed that the omission of the isotopes 146Nd, 148Nd, 150Nd, 242Cm, 243Cm, and 245 Cm from the MCNP cases has a negligible effect on system reactivity. The rationale for these isotopes being omitted is that the MCNP cross section libraries for these isotopes are not available, and their concentrations are very small (< 0.15 wt%). This assumption is used in Section 6. 3.7 It is assumed that using the Al material cross-section for Zn in the MCNP cases has a negligible impact on the results of criticality calculations. The basis for this assumption is that the neutronic characteristics for Zn and Al are sufficiently similar. The Zn neutron cross-section libraries are not available for MCNP. Also, the Zn material that is substituted only appears in Al6061 and is in trace amounts. This assumption is used in Section 5. 3.8 It is assumed that the length of each axial node in the core follow information for Limerick Unit 1 (Scaglione 2003) is 15.24 cm (6 in.). The basis for this assumption is the core follow information for LaSalle BWR (CRWMS M&O 1999, Table 3-4), which has a similar design with Limerick Unit 1 and the core follow information is reported for both reactors in the same (25 node) format. This assumption is used in Section 5.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 8 of 37
- 4. USE OF COMPUTER SOFTWARE AND MODELS 4.1 SOFTWARE APPROVED FOR QA WORK 4.1.1 SAS2H The SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN:
10129-4.4A-00), was used to perform the fuel assembly depletion calculations required for this evaluation. The software specifications are as follows:
- Program Name: SAS2H of the SCALE Modular Code System
- Version/Revision Number: Version 4.4A
- Status/Operating System: Qualified/HP-UX B.10.20
- Software Tracking Number: 10129-4.4A-00
- Computer Type: Hewlett Packard (HP) 9000 Series Workstations
- Computer Processing Unit number: 700887.
The input and output files for the various SAS2H calculations were documented in Attachments II and III to this calculation so that an independent repetition of the software use could be performed. The SAS2H code sequence of SCALE that was used was (1) appropriate for the application of commercial fuel assembly depletion, (2) used only within the range of validation documented in Users Manual for SCALE-4.4A (CRWMS M&O 2000a) and Validation Test Report (VTR) for SCALE-4.4A (CRWMS M&O 2000b), and (3) obtained from Software Configuration Management in accordance with appropriate procedures. 4.1.2 MCNP The baselined code MCNP, Version 4B2LV (CSCI: 30033-V4B2LV), was used to calculate the neutron multiplication factor for the various spent nuclear fuel compositions. The software specifications are as follows:
- Program Name: MCNP
- Version/Revision Number: Version 4B2LV
- Status/Operating System: Qualified/HP-UX B.10.20
- Computer Software Configuration Item Number: 30033-V4B2LV
- Computer Type: HP 9000 Series Workstations
- CPU number: 700887.
The input and output files for the various MCNP calculations are documented in Attachments II and III to this calculation so that an independent repetition of the software use may be performed. The MCNP software used was (1) appropriate for the application of multiplication factor calculations, (2) used only within the range of validation as documented throughout Software Qualification Report for MCNP Version 4B2, A General Monte Carlo N-Particle Transport Code (CRWMS M&O 1998) and MCNP-A General Monte Carlo N-Particle Transport Code (Briesmeister 1997), and (3) obtained from Software Configuration Management in accordance with appropriate procedures.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 9 of 37
- 5. CALCULATION This report evaluates the accuracy to which the SAS2H control module can predict the composition of spent nuclear fuel for eight different radiochemical assay (RCA) samples from assembly YJ1433 that operated at the Limerick Generation Station Unit 1 BWR.
5.1 CALCULATION METHOD The method of calculation is based upon the calculation of isotopic concentrations of irradiated fuel using the SAS2H sequence of the SCALE modular code system. All SAS2H inputs were set up to represent the assembly axial node where the measured fuel sample was located. 5.1.1 SAS2H Fuel Depletion Description The SAS2H control sequence accesses five calculation modules of the SCALE code system for performing fuel depletion and decay calculations. The five modules include BONAMI, NITAWL-II, XSDRNPM, COUPLE, and ORIGEN-S. Each of the modules has a specific purpose in the sequence to perform the fuel depletion and decay calculations. The following provides a brief description of what each module does with a more detailed description being provided in CRWMS M&O (2000a).
- BONAMI - applies the Bondarenko method of resonance self-shielding to nuclides for which Bondarenko data are available.
- NITAWL-II - performs Nordheim resonance self-shielding corrections for nuclides that have resonance parameter data available.
- XSDRNPM - performs a one-dimensional (1-D) neutron transport calculation on a specified geometry to facilitate production of cell-weighted cross sections for fuel depletion calculations.
- COUPLE - updates all cross section constants included on an ORIGEN-S working nuclear data library with data from the cell-weighted cross section library obtained from the XSDRNPM calculation. Additionally, the weighting spectrum produced by XSDRNPM is applied to update all nuclides in the ORIGEN-S working library, which were not included in the XSDRNPM calculation.
- ORIGEN-S - performs point depletion, buildup, and decay calculations for the specified assembly irradiation history. Additionally, can be run as a stand-alone case to provide isotopic concentrations at various decay times.
The SAS2H control module uses ORIGEN-S to perform a point depletion calculation for the fuel assembly section described in the SAS2H input file. The ORIGEN-S module uses cell-weighted cross sections based on 1-D transport calculations performed by XSDRNPM. One-dimensional transport calculations are performed on two models, Path A and Path B, to calculate energy dependent spatial neutron flux distributions necessary to perform cross section cell-weighting calculations. The Path A model is simply a unit cell of the fuel assembly lattice containing a fuel rod. In the Path A model, the fuel, cladding, and moderator are modeled explicitly. The only modification
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 10 of 37 required to develop the Path A model is the conversion of the fuel assembly's square lattice unit cell perimeter to a radial perimeter conserving moderator volume within the unit cell (exterior to the fuel rod cladding). This modification is performed automatically by the SAS2H control module. A 1-D transport calculation is performed on the Path A model for each energy group, and the spatial flux distributions for each energy group are used to calculate cell-weighted cross sections for the fuel. The Path B model is a larger representation of the assembly than the Path A model. The Path B model approximates spectral effects due to heterogeneity within the fuel assembly such as water gaps, burnable poison rods, control rods, or axial power shaping rods. The structure of the Path B model is based on a uniform distribution of non-fuel lattice cells. In reality, most fuel assemblies do not have uniformly distributed non-fuel lattice cells, but the approximation of uniformly distributed non-fuel lattice cells is considered acceptable within the fidelity of these calculations as documented in Section S2.2.3.1 of Volume 1, Rev. 6 in CRWMS M&O (2000a). The basic structure of the Path B model for the fuel assembly depletion calculations performed in this analysis included an inner region composed of a representation of the non-fuel assembly lattice cell. A region containing the homogenization of the Path A model surrounds the inner region in the Path B model. A final region representing the moderator in the assembly-to-assembly spacing surrounds the homogenized region in the Path B model. The size of each radial region that surrounds the inner region in the Path B model is determined by conserving both the fuel-to-moderator mass ratio and the fuel-to-absorber (burnable poison) mass ratio in the corresponding section of the fuel assembly. The cell-weighted cross sections from the Path A model are applied to the homogenized region during the Path B model transport calculations. New cell-weighted cross sections for each energy group are then developed using the unit cell spatial flux distribution results from the Path B model transport calculations. These cell-weighted cross sections are ultimately used in the point depletion calculations performed by ORIGEN-S to calculate the depleted fuel isotopic compositions in the corresponding fuel assembly. A detailed description of the calculations used to produce burnup-dependent cross sections by SAS2H is documented in Section S2.2.4 of Volume 1, Rev. 6 in CRWMS M&O (2000a). The Path B model for the fuel assembly configuration is provided to the SAS2H control module. The essential rule in deriving the zone radii is to maintain the relative volumes for all zones in the actual assembly (p. S2.2.5, CRWMS M&O 2000a). 5.2 SAMPLE DESCRIPTION Eight representative cross-sectional samples (approximately 1/2-inch thick) were obtained from three extended exposure fuel rods operated at the Limerick Unit 1 BWR to measure the concentrations of 37 selected isotopes (Reager 2003, pp. 1-1 and 2-1). The sampled fuel rods consisted of one full-length standard UO2 rod from fuel bundle lattice location D9; a gadolinia-bearing UO2 rod from location D8; and a part-length rod from location H5, all of which were retrieved from a GE11 9x9 fuel bundle assembly, identified as YJ1433. The 9x9 fuel rod array contains two water rods that span seven fuel rod positions, and eight part-length fuel rods (DOE 1996, p. 120). The samples were cut from various axial locations of the three fuel rods and were
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 11 of 37 dissolved and processed for the measurement of concentrations of the selected isotopes. Sample locations and basic characteristics are provided in Table 5-1. Table 5-1. Fuel Sample Characteristic Parameters Rod Location in Axial Positiona Initial Enrichment Measured Burnupb Assembly Lattice Sample Identifier (in. / cm) 235 (Wt% U) (MWd/MTU) D8-3D2B 82.30 / 209.04 D8 3.60 54,840 D8 - 3D2Cc 82.05 / 208.41 D8-4G3 130.45 / 331.34 D8 3.60 37,020 D8 - 4G4c 130.95 / 332.61 D9-1D2 30.75 / 78.11 D9 3.95 62,110 D9 - 1D3c 30.50 / 77.47 D9-2D2 62.25 / 158.12 D9 3.95 65,540 D9 - 2D3c 62.00 / 157.48 D9-4D4 102.20 / 259.59 D9 3.95 64,950 D9 - 4D3c 101.70 / 258.32 D9-4G1E1 123.06 / 312.57 D9 3.95 56,520 D9 - 4G1D1c 122.51 / 311.18 H5-3A1C 85.25 / 216.54 H5 3.95 57,915 H5 - 3A1Bc 85.00 / 215.90 H5-3A1G 88.90 / 225.81 H5 3.95 57,810 H5 - 3A1Ec 88.40 / 224.54 Source: Reager (2003, pp. 1-4 and 1-5). a NOTES: Axial position of each sample is measured in inches from the tip of bottom end plug of the fuel rod; cm value provided was converted from inches. b Measured burnup via 148Nd method. c Since the dissolution technique for samples assayed through the inductive coupled plasma mass spectroscopy (used for measurement of 95Mo, 99Tc, 101Ru, 103Rh, and 109Ag concentrations) was significantly different than the conditions used for the rest of the analyses, duplicate matched full cross-sectional fuel samples (approximately 1/4-inch thick) were excised from the fuel rods. The matched fuel samples were immediately adjacent or as close as possible to one another. No distinction will be made hereafter between the matched samples, i.e. the 95Mo, 99Tc, 101Ru, 103Rh, and 109Ag concentrations will be associated to the first sample in each pair in column Sample Identifier. For convenience, the rods will be identified hereafter by their lattice position within the fuel bundle. The dates of the measurements were used to calculate the amount of decay time associated with each sample. Reager (2003, pp. 4-3 through 4-7) indicated measurement dates of May 30, June 12 and July 15, 2002 for various samples. The date of July 15, 2002 was used as the reference date for all samples, and the measurements performed at other dates were corrected for decay to the reference date. This date corresponds to a downtime of 1510 days for all the samples investigated. The radiochemical assay measured isotopic concentrations for all eight samples from assembly YJ1433 are presented in Table 5-2.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 12 of 37 Table 5-2. Measured Concentrations for Assembly YJ1433 Samples Sample D8-3D2B D8-4G3 D9-1D2 D9-2D2 D9-4D4 D9-4G1E1 H5-3A1C H5-3A1G Average Burnup 54840 37020 62110 65540 64950 56520 57915 57810 (MWd/MTU) Isotope Concentration (mg/mg238U) 234 U 1.61E-04 1.94E-04 1.66E-04 1.63E-04 1.65E-04 1.85E-04 1.80E-04 1.87E-04 235 U 4.26E-03 8.72E-03 1.70E-03 2.19E-03 2.69E-03 4.31E-03 4.80E-03 4.88E-03 236 U 5.71E-03 5.18E-03 6.00E-03 6.01E-03 6.05E-03 6.01E-03 6.21E-03 6.21E-03 238 Pu 4.39E-04 2.55E-04 4.00E-04 5.08E-04 5.43E-04 4.45E-04 5.07E-04 5.08E-04 239 Pu 5.46E-03 5.52E-03 3.94E-03 4.77E-03 5.30E-03 5.44E-03 6.14E-03 6.18E-03 240 Pu 3.70E-03 2.90E-03 3.12E-03 3.50E-03 3.61E-03 3.35E-03 3.76E-03 3.77E-03 241 Pu 1.38E-03 1.14E-03 1.06E-03 1.31E-03 1.43E-03 1.39E-03 1.50E-03 1.50E-03 242 Pu 1.19E-03 6.33E-03 1.54E-03 1.62E-03 1.56E-04 1.23E-04 1.16E-03 1.14E-03 143 Nd 1.02E-03 9.19E-04 8.52E-04 9.85E-04 1.04E-03 1.04E-03 1.11E-03 1.11E-03 145 Nd 1.01E-03 7.99E-04 1.13E-03 1.17E-03 1.16E-03 1.06E-03 1.07E-03 1.07E-03 146 Nd 1.23E-03 8.52E-04 1.43E-03 1.52E-03 1.50E-03 1.28E-03 1.32E-03 1.31E-03 148 Nd 6.15E-04 4.44E-04 6.96E-04 7.36E-04 7.29E-04 6.37E-04 6.51E-04 6.49E-04 150 Nd 3.09E-04 2.16E-04 3.42E-04 3.68E-04 3.65E-04 3.15E-04 3.25E-04 3.23E-04 134 Cs 5.74E-05 3.35E-05 5.90E-05 7.17E-05 7.22E-05 5.66E-05 6.24E-05 6.21E-05 137 Cs 1.87E-03 1.35E-03 1.99E-03 2.17E-03 2.09E-03 1.78E-03 1.95E-03 1.94E-03 151 Eu 4.77E-07 4.32E-07 3.42E-07 4.04E-07 4.38E-07 4.52E-07 5.19E-07 5.26E-07 153 Eu 1.94E-04 1.41E-04 2.09E-04 2.21E-04 2.15E-04 2.01E-04 1.97E-04 1.97E-04 155 Eu 7.96E-06 5.36E-06 8.01E-06 9.01E-06 8.74E-06 8.00E-06 8.11E-06 8.10E-06 147 Sm 2.88E-04 2.58E-04 2.95E-04 2.96E-04 2.90E-04 2.99E-04 2.94E-04 2.95E-04 149 Sm 2.69E-06 2.91E-06 1.67E-06 2.35E-06 2.81E-06 3.08E-06 2.94E-06 2.95E-06 150 Sm 4.81E-04 3.34E-04 4.98E-04 5.51E-04 5.43E-04 4.90E-04 5.07E-04 5.04E-04 151 Sm 1.35E-05 1.23E-05 1.01E-05 1.24E-05 1.35E-05 1.35E-05 1.54E-05 1.53E-05 152 Sm 1.52E-04 1.20E-04 1.79E-04 1.76E-04 1.66E-04 1.58E-04 1.49E-04 1.48E-04 155 Gda 1.19E-05 9.68E-06 7.80E-06 8.78E-06 9.18E-06 7.79E-06 8.60E-06 8.09E-06 241 Am 3.89E-04 3.17E-04 2.83E-04 3.26E-04 3.81E-04 3.69E-04 4.08E-04 4.14E-04 241m Am 1.18E-06 1.16E-06 6.68E-07 8.18E-07 9.90E-07 1.06E-06 1.42E-06 1.46E-06 243 Am 3.03E-04 1.30E-04 3.62E-04 3.92E-04 4.19E-04 3.01E-04 2.96E-04 2.97E-04 237 Np 7.92E-04 5.40E-04 7.67E-04 9.00E-04 8.86E-04 8.22E-04 8.60E-04 8.65E-04 95 Mo 1.27E-03 9.94E-04 1.43E-03 1.42E-03 1.37E-03 1.27E-03 1.28E-03 1.28E-03 99 Tcb 1.27E-03 9.99E-04 1.39E-03 1.36E-03 1.40E-03 1.16E-03 1.22E-03 1.18E-03 101 Ru 1.28E-03 9.56E-04 1.50E-03 1.51E-03 1.47E-03 1.33E-03 1.32E-03 1.36E-03 103 Rh 7.45E-04 6.07E-04 7.22E-04 7.59E-04 8.00E-04 7.41E-04 7.54E-04 7.84E-04 109 Ag 1.61E-04 1.10E-04 1.46E-04 1.53E-04 1.55E-04 1.21E-04 1.34E-04 1.51E-04 242 Cm 4.91E-08 3.73E-08 3.75E-08 3.76E-08 5.28E-08 5.56E-08 4.81E-08 4.27E-08 243 Cm 1.15E-06 6.48E-07 8.49E-07 1.13E-06 1.37E-06 1.21E-06 1.24E-06 1.30E-06 244 Cm 1.56E-04 4.61E-05 1.81E-04 2.21E-04 2.43E-04 1.51E-04 1.62E-04 1.63E-04 245 Cm 1.28E-05 3.28E-06 9.50E-06 1.53E-05 1.90E-05 1.13E-05 1.53E-05 1.53E-05 246 Cm 2.94E-06 4.17E-07 3.46E-06 5.10E-05 5.54E-06 2.47E-06 3.10E-06 3.10E-06 Source: Reager (2003, pp. 4-2 through 4-9). a NOTE: Not including 155Gd from the decay of 155Eu. b Includes any 99Ru present in the sample solution.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 13 of 37 5.3 FUEL ASSEMBLY DESIGN AND OPERATING PARAMETERS The general fuel assembly design parameters are presented in Table 5-3. Table 5-3. YJ1433 Fuel Assembly Geometric and Material Information Fuel Assembly Data Lattice 9x9 Number of lattice positions occupied by water rods (DOE 1996, p. 120) 7 Number of fuel rods 74 Number of rods containing Gd2O3 9 Number of part-length fuel rods (DOE 1996, p. 120) 8 Rod pitch (cm) 1.440 Assembly pitch (cm) 15.240 Assembly channel material (see Assumption 3.1) Zircaloy-4 Fuel Rod Data Cladding outer diameter (cm) 1.1176 Cladding thickness (cm) 0.0711 Cladding inner diameter (cm) (= Outer diameter - 2*Thickness) 0.9754 Fill gas (DOE 1992, p. 2.2-3) Helium Cladding material Zircaloy-2 Fuel Pellet Data Diameter (cm) 0.9550 Pellet material UO2 Source: Nuclear Engineering International (1998, p. 63), unless otherwise noted. The nominal characteristics of the fuel rods from which the samples were cut are presented in Table 5-4. Table 5-4. Nominal Characteristics of Sampled Fuel Rods Fuel Rod As-Built 235U As-Built Gd2O3 Active Fuel Column As-Built Rod Length Average Exposure Number (Wt.%) (Wt.%) (in./cm) (in./cm) (GWd/MTU) D8a 3.60 5.00 138.0 / 350.5 160.97 / 408.9 43.5 D9b 3.95 0.00 146.0 / 370.8 160.97 / 408.9 54.2 H5 3.95 0.00 90.0 / 228.6 102.99 / 261.6 51.0 Source: Reager (2003, Table 1-2). a NOTES: D8: bottom 6 inches: UO2 (natural U), 112 g; top 132 inches: 95 wt% UO2 (3.60 wt% 235U) and 5 wt% Gd2O3, 2349g. b D9: bottom 6 inches: UO2 (natural U), 114 g; middle 132-inches: UO2 (3.95 wt% 235U), 2510g; top 8 inches UO2 (natural U), 152 g. There are eight part-length fuel rods like H5 in assembly YJ1433. Therefore, two main axial zones can be distinguished for assembly YJ1433:
- The lower zone, consisting of the lower 248.13 cm (97.69 in.) of the assembly fuel rods, as measured from the lower end plug tips of the rods. The length of this zone is the sum between the length of the active fuel column for the partial-length fuel rods (90 in. - see Table 5-4) and the remaining length of the fuel rod, from bottom end of active fuel
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 14 of 37 column to the end plug tip (7.69 in. - from Martinez 2001). This zone has 74 fuels rods (81 total lattice positions minus 7 positions occupied by water rods), of which 9 contain Gd2O3. Samples D8-3D2B, D8-3D2C, D9-1D2, D9-1D3, D9-2D2, D9-2D3, H5-3A1C, H5-3A1B, H5-3A1G, and H5-3A1E are axially located in this zone.
- The upper zone, which is between the top of the lower zone and the top end of the active fuel column of the standard rods. This zone has 66 fuels rods (81 total lattice positions minus 7 positions occupied by water rods, and minus 8 positions occupied by water above the top of part-length rods), of which 9 contain Gd2O3. Samples D8-4G3, D8-4G4, D9-4D4, D9-4D3, D9-4G1E1, and D9-4G1D1 are axially located in this zone.
5.3.1 SAS2H Material Specifications The material specification section defines the UO2 fresh fuel composition to which the SAS2H calculation pertains, along with the other materials necessary to describe the fuel assembly. The UO2 fresh fuel composition is characterized by the fuel density, fuel temperature, and weight percentages of 234U, 235U, and 238U. In SAS2H inputs, a number of additional isotopes are specified in trace amounts in the fresh fuel composition to assure that their buildup and decay is tracked during the depletion calculation. Table 5-5 contains a list of trace isotopes, which are specified as each having a concentration of 10-21 atoms/barn-cm in the fresh fuel composition. Table 5-5. Trace Isotopes Specified in Fresh Fuel Compositions 83 85 90 89 95 93 94 Kr Kr Sr Y Mo Zr Zr 95 94 95 99 103 105 101 Zr Nb Nb Tc Rh Rh Ru 106 105 108 109 124 126 131 Ru Pd Pd Ag Sb Sn Xe 132 135 136 134 135 137 136 Xe Xe Xe Cs Cs Cs Ba 139 144 141 143 143 144 145 La Ce Pr Pr Nd Nd Nd 146 147 148 150 147 148 149 Nd Nd Nd Nd Pm Pm Pm 147 148 149 150 151 152 154 Sm Sm Sm Sm Sm Sm Gd 155 157 158 160 151 153 154 Gd Gd Gd Gd Eu Eu Eu 155 232 233 Eu U U The wt% ranges for Zircaloy-2 and Zircaloy-4 compositions as given in MO9906RIB00048.000 and the values selected from these ranges for the SAS2H inputs are presented in Table 5-6. Table 5-6. Zircaloy-2 and Zircaloy-4 Compositions Element Wt% in Zircaloy-2a SAS2H Wt% in Zircaloy-2 Wt% in Zircaloy-4a SAS2H Wt% in Zircaloy-4 Cr 0.05-0.15 0.10 0.07-0.13 0.10 Fe 0.07-0.20 0.135 0.18-0.24 0.21 O 0.09-0.16 0.125 0.09-0.16 0.125 Sn 1.20-1.70 1.45 1.20-1.70 1.45 Ni 0.03-0.08 0.055 - - Fe+Cr 0.18-0.38 - 0.28-0.37 - Zr Remainder 98.135 Remainder 98.115 Densitya 6.55 g/cm3 6.56 g/cm3 a Source: MO9906RIB00048.000.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 15 of 37 In the SAS2H Path A model, the moderator density and temperature must be provided, and both remain unchanged unless new values are added to the input irradiation history. The Limerick Unit 1 core follow report (Scaglione 2003) contains axial profile information for core averaged moderator void fraction, power, and exposure, and for YJ1433 assembly channel averaged power and exposure. The axial profile information for each of these physical quantities is provided as a sequence of 25 values, one for each axial node, which is a slice 15.24-cm- (6-in.-) thick of the active core (see Assumption 3.8). The lower end of the first node coincides with the lower end of the active fuel column of the standard fuel pin, which is 3.97 cm (1.563 in.) from the bottom end plug tip (Martinez 2001). The axial profile information in Scaglione (2003) is provided for 20 accounting cases for core operating cycle 5, 35 for cycle 6, and 32 for cycle 7. This information can be used to estimate the axial profile of the moderator density inside assembly YJ1433 channel, for each exposure accounting case. Therefore, it is possible to develop a more realistic Path A model than the simplified model that uses a constant moderator density for the entire residence time of the assembly in the core. Since the moderator flows upward through the channel, the moderator density at any location is dependent not only on the power at that location but also on the power at lower locations. A correlation between the core averaged node moderator density was developed as a function of the node integrated relative core power. The node integrated relative core power is defined as the sum of the powers for that node and all the nodes below: i Pi = p k (Eq. 1) 1 where pk = relative core power for node k (normalized to 1); Pi = integrated core relative power for node i. The variation of core averaged node moderator density with node integrated relative core power is well fit as a fifth degree polynomial (BSC 2001, Section 5.2.7): 5 i = C j Pi j (Eq. 2) j= 0 The coefficients Cj are calculated using the least squares method (Walpole et al. 1998, p. 411). As explained below, these coefficients are used to calculate the node averaged moderator density inside the channel containing assembly YJ1433. The node integrated relative core power is defined as the sum of the powers for that node and all the nodes below: i P1433i = p1433k (Eq. 3) 1
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 16 of 37 where p1433k = relative channel (containing assembly YJ1433) power for node k (normalized to 1); P1433i = integrated channel (containing assembly YJ1433) relative power for node i. The node averaged moderator density in the channel containing assembly YJ1433 is calculated using the coefficients calculated via the least squares method in the following equation: 5 1433i = C j P1433 j i (Eq. 4) j= 0 The isotopic composition of 234U in fresh fuel was calculated using the following equation based on Assumption 3.2: 234 U wt% = 0.009
- 235U wt% (Eq. 5)
The 238U wt% is then the remainder from the total of 100 wt%: 238 U wt% = 100 wt% - 234U wt% - 235U wt% (Eq. 6) The fresh fuel densities and compositions used in this calculation are specified in Table 5-7. Table 5-7. SAS2H Compositions Specific to Assembly YJ1433 Densityb (g/cm3) Concentrationc (wt%) SAS2H Rod D8 Rod D9 and H5 SAS2H Rod D8 Rod D9 and H5 Description Mixture # samples samples Identifier samples samples 92234 0.03240 0.03555 Fuel 1 9.37605 10.45128 92235 3.60000a 3.95000a 92238 96.36760 96.01445 UO2-Gd2O3 64000 a 4 9.37605 10.45128 5.0 0a Rod 16000 a Source: Reager (2003, Table 1-2). b NOTES: The fuel density was calculated using the fuel column masses in pins D8 and D9 given in Reager (2003, Table 1-2), and by expanding the rod UO2 loading to fill the cladding-pellet gap. c The 234U and 238U wt.% were calculated using Equations 5 and 6, and the 235U wt.% given in Reager (2003, Table 1-2). 5.3.2 SAS2H Operating History Specifications The core follow report (Scaglione 2003) provides cumulative burnup information for assembly YJ1433. The average burnups for each axial node of the assembly are provided for 20 accounting cases for core operating cycle 5, 35 for cycle 6, and 32 for cycle 7. For each sample is known only the final burnup (see Table 5-1), which was measured through radiochemical assay. Sample cumulative burnup histories were estimated considering each sample was burned at the same relative rate as the full assembly. Thus, sample burnups per accounting case were obtained by normalizing the assembly burnup history and multiplying by the final sample burnup.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 17 of 37 The axial power and burnup profile information in Scaglione (2003) was used to calculate the average values for assembly node power, moderator density, and temperatures of fuel, cladding and moderator in 37 irradiation steps, with lengths between 35 and 81 days. Irradiation steps that are less than 90 days are considered sufficiently short to have a negligible impact on SAS2H depletion results. The fuel temperature corresponding to the nodal power was calculated from the nodal power in agreement with assumptions 3.4 and 3.5, by using equations 7 through 12, and the physical constants listed in Table 5-8. Table 5-8. Constants for Fuel Temperature Calculations Constant Value used Reference (Cladding Inner Diameter + Pellet Outer Cladding-pellet gap mean diameter, DG (cm) 0.965 Diameter)/2 (diameters taken from Table 5.3) Effective gap thickness (cm) 0.0112 Calculated using Equation 10 4 Stefan-Boltzmann constant (kWcm K ) 5.67E-15 Parrington et al. 1996, p. 59 Fuel conductivity (kWcm-1K-1) 3.17E-05 Hagrman et al. 1981, p. 29 Cladding conductivity (kWcm-1K-1) 1.64E-04 Hagrman et al. 1981, p. 219 Fuel emissivity 0.7993 Hagrman et al. 1981, p. 48 Cladding emissivity 0.325 Hagrman et al. 1981, p. 230 The linear heat generation rate ( q& ) for a single fuel rod is calculated using Equation 7. Pnode K 235 q& = (Eq. 7) N rods L node where Pnode = average assembly power in node (kW); K235 = ratio between pin and assembly 235U wt.% (used to correct the power in pin); Nrods = the number of fuel rods in lattice; Lnode = the active length of the node (cm). The heat flux at the surface of the fuel pellet ( &q& ) is calculated using Equation 8. q&
&q& = (Eq. 8)
DP where DP is the diameter of the fuel pellet (cm). The temperature change across the cladding (TC) is calculated using Equation 9 (Todreas and Kazimi 1990, p. 336).
&q& C TC = (Eq. 9) kC
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 18 of 37 where C = the thickness of the cladding (cm); kC = the cladding conductivity (kWcm-1K-1). The effective gap thickness (eff) is estimated using Equation 10 (Todreas and Kazimi 1990, p. 334). eff = G + jump1 + jump 2 (Eq. 10) where G is the geometrical gap thickness (cm), and jump1 and jump2 are corrections for temperature discontinuities near the surfaces (at atmospheric pressure, the sum of the last two terms is known to be 10-3 cm for helium; Todreas and Kazimi 1990, p. 334). The gap thermal conductance (hG) is calculated as follows (Todreas and Kazimi 1990, p. 334): kG TS3 hG = + (Eq. 11) eff 1 1
+ 1 F C where TS = the fuel pellet surface temperature (K);
kG = the thermal conductivity of the gas in the gap (kWcm-1K-1);
= the Stefan-Boltzmann constant (5.6710-15 kWcm-2K-4);
F and C = the emissivities of the fuel and cladding, respectively. The thermal conductivity of the gas in the gap (helium), kG is dependent on temperature as follows (Todreas and Kazimi 1990, p. 334): k G = 15.8 106 T 0.79 (Eq. 12) where T (K) is the gas temperature that was approximated with the average temperature across the gap (Tavgg, which is calculated using Equation 15). For the fuel/cladding gap without gap closure, the temperature jump is given as follows (Todreas and Kazimi 1990, p. 336): q& TG = (Eq. 13) DG h G where DG = the gap mean diameter (cm); hG = the gap conductance (kWcm-2K-1; calculated using Equation 11). The bulk temperature in the coolant/moderator (Tbulk) was considered approximately equal to the core outlet temperature, which was calculated by adding 10 K to the core inlet temperature. The
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 19 of 37 value of 10 K (or 10 °C) is used because it is equal to the difference between the core outlet (288 °C) and inlet (278 °C) temperatures for the Grand Gulf Unit 1 BWR, which has a very similar design to that of Limerick Unit 1 BWR (Nuclear Engineering International 1998, pp. 54 and 63). Assuming excellent heat transfer due to boiling (see Assumption 3.4) and given the bulk temperature (Tbulk) in the coolant, the surface temperature of the fuel pellet may be calculated as follows: TS = Tbulk + TC + TG (Eq. 14) The average temperature across the gap is calculated as follows: TG Tavgg = Tbulk + TC + (Eq. 15) 2 The surface temperature of the fuel pellet and the linear heat generation rate were used to determine the centerline temperature of the fuel as shown in Equation 16 (Todreas and Kazimi 1990, p. 336): q& TCL = + TS (Eq. 16) 4 kf where TCL = fuel centerline temperature (maximum fuel temperature; K); kf = fuel thermal conductivity (3.17E-05 kWcm-1K-1, which is based on 93.4 % of theoretical density and an average temperature of 1071 K; from Hagrman et al. 1981, p. 29). The average temperature across the cladding is calculated as follows: TC Tavgc = TBulk + (Eq. 17) 2 The fuel, cladding, and moderator average temperatures for each sample and each irradiation step are presented in Table 5-9. The fuel conductivity was considered to be constant, thus the average fuel temperature (Tavgf) is the simple average of the pellet surface and centerline temperatures. The source of information in Table 5-9 is spreadsheet LGS1RCA.xls in Attachment III (in the worksheets that have identical names with the samples). The specific columns in the worksheets from where the information is retrieved are presented in the next eight paragraphs. The power, density and temperature values used in the 37 power steps for SAS2H input are highlighted in yellow in the worksheets. In Table 5-9 the column headings are the same as the keywords used in the power history data block of the SAS2H input:
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 20 of 37 power - nodal power for a given irradiation step (MW/node; in column with the header Linear Heat Gen. Rate). It was calculated by dividing the YJ1433 assembly exposure for the node where the sample is located as given in Scaglione (2003) in MWD/ST for each time step to the length of the time step (days) and multiplied with a correction factor with the purpose of obtaining the same burnup as the sample (see also first paragraph in this section). burn - duration of the irradiation step (days; in the second column with the header Days). down - time at zero power (days; in cells D100, D191, and D281). h2of - node averaged fraction of first power step moderator density (in column with the header H2O fraction). The method of calculating the moderator density is explained in Section 5.3.1. tmpf - node averaged fuel temperature for the irradiation step (K; in column with the header Fuel Tavgf). tmpc - node averaged fuel cladding temperature for the irradiation step (K; was taken from column with the header Cladding Tavgc). tmpm - node averaged moderator temperature for the irradiation step (K, was taken from column with the header Moderator Tbulk). Equations 7 through 17 were used to calculate the average temperatures for fuel, cladding and moderator. Because the pellet surface temperature and the average temperature in the fuel-cladding gap were unknowns in a system of two transcendental equations, these temperatures were determined by changing the given input values (in columns with the headers Input Value for TS, and Input Value for Tavgg) until they matched the calculated values.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 21 of 37 Table 5-9. Operating History Information for Assembly YJ1433 Node 14 (D8-3D2B) Node 22 (D8-4G3) Node 5 (D9-1D2) burn down power h2of tmpf tmpc tmpm power h2of tmpf tmpc tmpm power h2of tmpf tmpc tmpm 73.0 0 0.19348 1.000 1059 574 561 0.08331 1.000 854 568 561 0.19255 1.000 1026 573 561 45.5 0 0.19718 0.877 1068 575 561 0.08553 0.902 862 569 561 0.28620 0.939 1233 579 561 45.5 0 0.19718 0.877 1068 575 561 0.08553 0.902 862 569 561 0.28620 0.939 1233 579 561 45.5 0 0.20425 0.880 1085 575 561 0.10600 0.896 929 570 561 0.28423 0.900 1228 579 561 45.5 0 0.20425 0.880 1085 575 561 0.10600 0.896 929 570 561 0.28423 0.900 1228 579 561 45.5 0 0.20348 0.978 1083 575 561 0.12238 0.905 982 572 561 0.25907 0.974 1174 577 561 45.5 0 0.20348 0.978 1083 575 561 0.12238 0.905 982 572 561 0.25907 0.974 1174 577 561 55.0 0 0.23563 0.878 1160 578 562 0.12725 1.065 998 573 562 0.31797 0.927 1301 582 562 53.0 0 0.21628 0.934 1114 577 562 0.12981 1.027 1007 574 562 0.23395 0.965 1120 577 562 50.0 0 0.24685 1.073 1183 575 558 0.16590 0.916 1118 573 558 0.23235 1.035 1113 573 558 67.0 38.0 0.20453 1.272 1086 575 561 0.16060 0.827 1103 575 561 0.13960 1.073 905 570 561 57.0 0 0.26255 0.696 1220 577 559 0.12149 1.774 978 570 559 0.33150 0.848 1328 580 559 65.0 0 0.24847 0.678 1189 577 560 0.12530 1.511 991 572 560 0.36368 0.805 1396 584 560 42.5 0 0.23814 0.725 1164 576 559 0.11584 1.246 960 570 559 0.36585 0.835 1400 583 559 42.5 0 0.23814 0.725 1164 576 559 0.11584 1.246 960 570 559 0.36585 0.835 1400 583 559 37.0 0 0.23024 0.779 1147 577 562 0.13369 1.140 1019 574 562 0.29116 0.845 1244 580 562 67.0 0 0.23010 0.793 1146 576 560 0.13108 1.041 1010 572 560 0.30643 0.844 1276 580 560 37.0 0 0.20482 0.819 1086 574 560 0.15234 0.910 1077 573 560 0.26226 0.875 1180 576 560 67.0 0 0.22097 0.889 1124 575 560 0.16937 0.835 1130 575 560 0.23495 0.971 1120 575 560 35.0 0 0.19817 0.847 1070 574 560 0.11503 0.850 958 571 560 0.22000 0.932 1088 574 560 42.0 0 0.21758 0.947 1116 575 560 0.13324 0.900 1016 572 560 0.21928 0.984 1086 574 560 43.0 0 0.14627 0.955 945 572 562 0.10259 0.887 918 571 562 0.13279 1.001 890 570 562 43.0 0 0.14627 0.955 945 572 562 0.10259 0.887 918 571 562 0.13279 1.001 890 570 562 60.0 0 0.21285 1.078 1105 574 560 0.16603 0.966 1120 575 560 0.17655 1.040 989 571 560 45.0 27.0 0.19883 1.293 1072 574 560 0.16373 1.161 1113 575 560 0.13438 1.077 893 569 560 57.0 0 0.16851 0.926 1001 575 563 0.06781 0.951 804 569 563 0.15984 1.007 954 573 563 78.0 0 0.19549 0.873 1065 575 562 0.07698 0.906 834 569 562 0.20550 0.977 1057 575 562 60.5 0 0.20135 0.894 1079 576 562 0.09416 0.876 891 571 562 0.20948 0.948 1066 576 562 60.5 0 0.20135 0.894 1079 576 562 0.09416 0.876 891 571 562 0.20948 0.948 1066 576 562 71.0 0 0.19366 0.952 1061 576 562 0.11309 0.921 953 573 562 0.19830 1.001 1041 575 562 74.0 0 0.20782 0.961 1095 576 562 0.10944 0.935 941 572 562 0.18613 1.000 1013 574 562 71.0 0 0.18913 1.111 1051 576 563 0.14128 0.992 1044 576 563 0.16350 1.026 962 574 563 81.0 0 0.15965 1.157 979 574 563 0.13612 1.066 1028 575 563 0.17005 1.021 977 574 563 41.0 0 0.15220 1.101 960 572 562 0.13692 1.046 1029 574 562 0.18649 0.994 1014 574 562 48.0 0 0.16887 1.000 1002 575 563 0.12618 1.008 996 575 563 0.23810 0.920 1130 579 563 48.0 0 0.16091 1.073 983 574 563 0.12809 1.060 1003 575 563 0.18756 0.965 1017 575 563 63.0 1510 0.12927 1.307 902 569 560 0.10709 1.247 932 569 560 0.13378 1.052 891 568 560
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 22 of 37 Table 5-9. Operating History Information for Assembly YJ1433 (continued) Node 11 (D9-2D2) Node 17 (D9-4D4) Node 21 (D9-4G1E1) burn down power h2of tmpf tmpc tmpm power h2of tmpf tmpc tmpm power h2of tmpf tmpc tmpm 73.0 0 0.23202 1.000 1107 575 561 0.18271 1.000 1062 574 561 0.13137 1.000 930 570 561 45.5 0 0.26529 0.853 1180 578 561 0.18458 0.882 1067 575 561 0.13469 0.891 939 571 561 45.5 0 0.26529 0.853 1180 578 561 0.18458 0.882 1067 575 561 0.13469 0.891 939 571 561 45.5 0 0.25452 0.848 1156 577 561 0.21496 0.874 1143 576 561 0.16663 0.876 1022 573 561 45.5 0 0.25452 0.848 1156 577 561 0.21496 0.874 1143 576 561 0.16663 0.876 1022 573 561 45.5 0 0.23305 0.977 1110 576 561 0.24941 0.946 1228 579 561 0.19459 0.898 1093 575 561 45.5 0 0.23305 0.977 1110 576 561 0.24941 0.946 1228 579 561 0.19459 0.898 1093 575 561 55.0 0 0.29177 0.872 1237 580 562 0.26319 0.844 1263 581 562 0.20343 0.959 1116 577 562 53.0 0 0.25663 0.933 1162 578 562 0.24836 0.893 1226 580 562 0.20660 0.937 1124 577 562 50.0 0 0.28952 1.082 1229 576 558 0.28455 1.023 1312 579 558 0.25913 0.917 1252 577 558 67.0 38.0 0.23129 1.268 1106 575 561 0.23933 1.228 1204 578 561 0.24137 0.978 1210 579 561 57.0 0 0.33564 0.716 1327 580 559 0.25400 0.729 1238 578 559 0.19225 1.436 1086 573 559 65.0 0 0.32068 0.684 1296 581 560 0.24489 0.700 1217 578 560 0.19441 1.235 1092 574 560 42.5 0 0.30160 0.726 1256 578 559 0.23708 0.699 1197 576 559 0.18062 1.009 1056 572 559 42.5 0 0.30160 0.726 1256 578 559 0.23708 0.699 1197 576 559 0.18062 1.009 1056 572 559 37.0 0 0.27503 0.766 1201 579 562 0.25360 0.759 1239 580 562 0.20588 0.990 1122 577 562 67.0 0 0.27154 0.773 1193 578 560 0.26351 0.768 1263 580 560 0.20480 0.922 1119 575 560 37.0 0 0.23440 0.803 1112 574 560 0.23457 0.796 1191 577 560 0.23120 0.823 1184 577 560 67.0 0 0.24842 0.905 1142 575 560 0.25440 0.858 1240 578 560 0.25640 0.798 1246 578 560 35.0 0 0.23686 0.844 1118 575 560 0.20907 0.824 1128 576 560 0.17684 0.815 1048 573 560 42.0 0 0.25812 0.951 1164 576 560 0.22919 0.930 1178 577 560 0.20237 0.891 1112 575 560 43.0 0 0.16652 0.969 963 572 562 0.15925 0.953 1002 573 562 0.15184 0.907 984 573 562 43.0 0 0.16652 0.969 963 572 562 0.15925 0.953 1002 573 562 0.15184 0.907 984 573 562 60.0 0 0.24491 1.098 1135 575 560 0.23439 1.049 1191 577 560 0.23936 0.986 1204 577 560 45.0 27.0 0.22816 1.304 1099 575 560 0.21826 1.241 1151 576 560 0.23172 1.188 1186 577 560 57.0 0 0.21049 0.942 1062 576 563 0.15473 0.931 992 574 563 0.10735 0.948 868 571 563 78.0 0 0.25345 0.873 1155 578 562 0.17428 0.887 1041 575 562 0.12128 0.904 904 571 562 60.5 0 0.23769 0.888 1121 577 562 0.19138 0.896 1085 576 562 0.14585 0.882 969 573 562 60.5 0 0.23769 0.888 1121 577 562 0.19138 0.896 1085 576 562 0.14585 0.882 969 573 562 71.0 0 0.23943 0.972 1125 578 562 0.19113 0.943 1084 576 562 0.16984 0.928 1031 575 562 74.0 0 0.25311 0.983 1154 578 562 0.19808 0.955 1102 577 562 0.16665 0.942 1023 574 562 71.0 0 0.18066 1.119 996 575 563 0.21880 1.066 1154 579 563 0.21403 1.009 1144 579 563 81.0 0 0.17242 1.143 977 574 563 0.17623 1.138 1047 576 563 0.20274 1.086 1115 578 563 41.0 0 0.18265 1.069 999 573 562 0.17503 1.104 1043 574 562 0.20369 1.062 1117 577 562 48.0 0 0.19833 0.957 1035 576 563 0.17902 1.025 1054 576 563 0.18537 1.018 1072 577 563 48.0 0 0.18429 1.038 1004 575 563 0.17331 1.087 1040 576 563 0.18472 1.074 1070 577 563 63.0 1510 0.15039 1.256 925 569 560 0.13888 1.304 948 570 560 0.15201 1.264 983 571 560
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 23 of 37 Table 5-9. Operating History Information for Assembly YJ1433 (continued) Node 14 (H5-3A1C) Node 15 (H5-3A1G) burn down power h2of tmpf tmpc tmpm power h2of tmpf tmpc Tmpm 73.0 0 0.20433 1.000 1104 575 561 0.19860 1.000 1087 575 561 45.5 0 0.20823 0.877 1114 576 561 0.19891 0.881 1089 575 561 45.5 0 0.20823 0.877 1114 576 561 0.19891 0.881 1089 575 561 45.5 0 0.21570 0.880 1132 576 561 0.21298 0.882 1123 576 561 45.5 0 0.21570 0.880 1132 576 561 0.21298 0.882 1123 576 561 45.5 0 0.21489 0.978 1130 576 561 0.21974 0.972 1139 576 561 45.5 0 0.21489 0.978 1130 576 561 0.21974 0.972 1139 576 561 55.0 0 0.24884 0.878 1213 580 562 0.24720 0.871 1206 579 562 53.0 0 0.22841 0.934 1164 578 562 0.22941 0.927 1163 578 562 50.0 0 0.26069 1.073 1239 576 558 0.26297 1.062 1241 577 558 67.0 38.0 0.21600 1.272 1133 576 561 0.21927 1.261 1138 576 561 57.0 0 0.27727 0.696 1279 579 559 0.26841 0.689 1255 578 559 65.0 0 0.26241 0.678 1245 579 560 0.25473 0.673 1223 578 560 42.5 0 0.25149 0.725 1218 577 559 0.24569 0.715 1201 577 559 42.5 0 0.25149 0.725 1218 577 559 0.24569 0.715 1201 577 559 37.0 0 0.24315 0.779 1199 579 562 0.25029 0.772 1213 579 562 67.0 0 0.24300 0.793 1198 578 560 0.25549 0.788 1225 579 560 37.0 0 0.21630 0.819 1133 575 560 0.21936 0.817 1137 575 560 67.0 0 0.23336 0.889 1174 576 560 0.23707 0.883 1180 577 560 35.0 0 0.20928 0.847 1116 575 560 0.20801 0.844 1110 575 560 42.0 0 0.22978 0.947 1166 577 560 0.22823 0.944 1159 576 560 43.0 0 0.15448 0.955 980 573 562 0.15535 0.955 980 573 562 43.0 0 0.15448 0.955 980 573 562 0.15535 0.955 980 573 562 60.0 0 0.22478 1.078 1153 576 560 0.22645 1.069 1155 576 560 45.0 27.0 0.20997 1.293 1118 575 560 0.21121 1.275 1118 575 560 57.0 0 0.17795 0.926 1041 576 563 0.17272 0.926 1025 575 563 78.0 0 0.20645 0.873 1110 577 562 0.19865 0.877 1089 576 562 60.5 0 0.21264 0.894 1126 577 562 0.20975 0.895 1116 577 562 60.5 0 0.21264 0.894 1126 577 562 0.20975 0.895 1116 577 562 71.0 0 0.20451 0.952 1106 577 562 0.20211 0.947 1097 577 562 74.0 0 0.21947 0.961 1142 578 562 0.21509 0.958 1129 577 562 71.0 0 0.19974 1.111 1095 577 563 0.21091 1.095 1120 578 563 81.0 0 0.16860 1.157 1017 575 563 0.17305 1.152 1026 575 563 41.0 0 0.16073 1.101 996 573 562 0.16411 1.105 1003 573 562 48.0 0 0.17834 1.000 1042 576 563 0.18076 1.010 1045 576 563 48.0 0 0.16993 1.073 1021 576 563 0.17298 1.079 1026 576 563 63.0 1510 0.13652 1.307 933 570 560 0.13837 1.310 936 570 560 5.4 SAS2H PATH B REPRESENTATIONS The Path B model for the fuel assembly configuration is provided to the SAS2H control module. The primary concern in the development of the Path B model for BWR assemblies is the conservation of the fuel-to-moderator and the fuel-to-absorber mass ratios. The Path B model used in this evaluation is for a GE11 9x9 fuel assembly with 9 rods that contain gadolinium as a burnable neutron absorber. Figure 5-1 illustrates the general diagram of the Path B model, where Ri (i = 1,,6) refer to the radial dimensions.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 24 of 37 Homogenized Fuel Region Fuel-Gadolinia Region In-channel Moderator Bypass Moderator Cladding Channel R1 R2 R3 R4 R5 R6 Figure 5-1. SAS2H General Path B Model used for Assembly YJ1433 At any time during assembly YJ1433 residence inside Limerick Unit 1 BWR core, there is no control blade inserted fully or partially inside the blade position adjacent to the assembly (Scaglione 2003). Therefore no control blade region is represented in the Path B model. SAS2H code sets the fuel composition to be uniform throughout the fuel and burnable absorber region (CRWMS M&O 2000a, p. S2.5.15). In the SAS2H representation of the fuel assembly node, the input value for the U fuel composition that most accurately represents the node assembly is the node-averaged U composition. However, for the purpose of this calculation the most accurate sample representation is needed. Therefore, the input value used for the U fuel composition in each case was the sample composition. Equations 18 through 23 are used to calculate the Path B zone radii. For the first zone, which represents the burnable absorber rod (UO2-Gd2O3 mixture), the pellet diameter was increased to the inner cladding diameter and the material density was adjusted (smeared) accordingly. Therefore, the radius of the first zone is: R1 = (Fuel Rod Cladding Inner Diameter)/2 = 0.4877 cm. (Eq. 18) The second zone represents the fuel rod cladding. Its radius is: R2 = (Fuel Rod Cladding Outer Diameter)/2 = 0.5588 cm. (Eq. 19) The third zone represents the moderator in the fuel rod lattice cell. Its radius is: rodpitch R3 = = 0.81243 cm (Eq. 20) where rodpitch is the rod pitch in the 9x9 lattice. The radii values for the first three zones are applicable to all eight samples. The fourth zone represents the homogenized fuel zone. Its radius is:
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 25 of 37
# fuelrods rodpitch 2 R 4 = R 32 + 1 (Eq. 21)
# gdrods where R3 = the radius of the third Path B zone
#fuelrods = the total number of fuel rods in assembly (including those that contain Gd2O3)
#gdrods = the number of rods in assembly that contain Gd2O3.
As explained in Section 5.3, the fuel assembly has two axial zones with different numbers of fuel rods: 74 in the lower zone, and 66 in the upper zone. Therefore, Equation 21 gives two different results for R4, 2.32960 cm for the lower zone, and 2.20008 cm for the upper zone. The fifth zone represents the assembly channel zone. Its radius is: outwidth 2 inwidth 2 R 5 = R 24 + (Eq. 22)
# gdrods where R4 = the radius of the fourth Path B zone outwidth = the outer width of the assembly channel inwidth = the inner width of the assembly channel.
R5 is 2.45181 cm for the lower zone, and 2.32909 cm for the upper zone. The sixth zone represents the moderator outside the assembly channel zone, but inside the assembly lattice cell. Its radius is: assemblypitch R6 = (Eq. 23)
# gdrods where assemblypitch is the assembly pitch in the core zone lattice.
Equation 23 is independent of the axial zone of the assembly, therefore R6 is 2.86608 cm for both zones. 5.5 MCNP SPECIFICATIONS In order to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on system reactivity, MCNP calculations were performed to calculate the multiplication factor (keff) that results from using the different sets of isotopic concentrations and provide a comparison in terms of keff. The axial nodes from which the samples came from were used to represent fuel assemblies in a flooded waste package configuration. The waste package design parameters used in the MCNP
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 26 of 37 representations are illustrated in Attachment I. Axially reflective boundary conditions were used for each representation. The general assembly design parameters were presented in Table 5-3. The spent fuel isotopes used in the MCNP cases correspond to those from the SAS2H calculations and the measured sample isotopes. Isotopes were extracted from the SAS2H outputs and measured results, and then combined with the initial oxygen mass and renormalized to the total mass in terms of weight percents. Isotopes 146Nd, 148Nd, 150Nd, 242Cm, 243Cm, 245Cm, and 246Cm were omitted from the MCNP cases as they have a negligible effect on system reactivity (see Assumption 3.6). The values from the SAS2H calculations are given in units of mols, which were converted to units of grams using Equation 24. In order to keep changes in reactivity limited to variations from isotopic concentrations, for the MCNP density input was used the value calculated for the fresh fuel. Each depleted fuel composition is listed in the MCNP input files contained in Attachment III in terms of ZAIDs and weight percents, and can be verified by visual inspection from the SAS2H outputs along with the equation provided. The SAS2H output files for each calculation are contained on a compact disc attachment (Attachment III). Massi = (Mols Isotopei)
- Ai (Eq. 24) where i is the particular isotope and Ai is the atomic mass value (from Audi and Wapstra 1995).
The outer barrier of the waste package was represented as SB-575 N06022 as described in Table 5-10. The inner barrier was represented as SA-240 S31600, which is nuclear grade 316 stainless steel (SS) with tightened control on carbon and nitrogen content (ASM International 1987, p. 931, and ASME 1998, Section II, SA-240, Table 1) as described in Table 5-11. The fuel basket plates were represented as Neutronit A978 with 1.62 wt% boron as described in Table 5-12, and the thermal shunts were represented as aluminum 6061 as described in Table 5-13. The basket side and corner guides were represented as Grade 70 A 516 carbon steel as described in Table 5-14. The basket stiffeners were represented as water since they are not solid over the length of the basket. The chromium, nickel, and iron elemental weight percents obtained from the references were expanded into their constituent natural isotopic weight percents for use in MCNP. This expansion was performed by: (1) calculating a natural weight fraction of each isotope in the elemental state, and (2) multiplying the elemental weight percent in the material of interest by the natural weight fraction of the isotope in the elemental state to obtain the weight percent of the isotope in the material of interest. This process is described mathematically in Equations 25 and 26. The atomic mass values and atom percent of natural element values for these calculations are from Parrington et al (1996). Weight Fraction (Atomic Mass of Isotope "i")(At% of Isotope "i" in Natural Element ) of Isotope " i" in the = I Natural Element (Atomic Mass of Isotope"i")(At% of Isotope "i" in Natural Element ) i =1 (Eq. 25) where I the total number of isotopes in the natural element
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 27 of 37 Weight Fraction Wt % of Isotope "i" in Reference Wt % of
= of Isotope "i" in the Material Composition Natural Element Element in Material Composition (Eq. 26)
Table 5-10. Material Specifications for SB-575 N06022 Element/Isotope ZAID Wt% Element/Isotope ZAID Wt% 59 C-nat 6000.50c 0.0150 Co 27059.50c 2.5000 55 182 Mn 25055.50c 0.5000 W 74182.55c 0.7877 183 Si-nat 14000.50c 0.0800 W 74183.55c 0.4278 50 184 Cr 24050.60c 0.8879 W 74184.55c 0.9209 52 186 Cr 24052.60c 17.7863 W 74186.55c 0.8636 53 Cr 24053.60c 2.0554 V 23000.50c 0.3500 54 54 Cr 24054.60c 0.5202 Fe 26054.60c 0.2260 58 56 Ni 28058.60c 36.8024 Fe 26056.60c 3.6759 60 57 Ni 28060.60c 14.6621 Fe 26057.60c 0.0865 61 58 Ni 28061.60c 0.6481 Fe 26058.60c 0.0116 62 32 Ni 28062.60c 2.0975 S 16032.50c 0.0200 64 31 Ni 28064.60c 0.5547 P 15031.50c 0.0200 Mo-nat 42000.50c 13.5000 Density = 8.69 g/cm3 Source: MO0003RIB00071.000. Table 5-11. Material Specifications for SS316NG Element/Isotope ZAID Wt% Element/Isotope ZAID Wt% 54 C-nat 6000.50c 0.0200 Fe 26054.60c 3.6911 14 56 N 7014.50c 0.0800 Fe 26056.60c 60.0322 57 Si-nat 14000.50c 1.0000 Fe 26057.60c 1.4119 31 58 P 15031.50c 0.0450 Fe 26058.60c 0.1897 32 58 S 16032.50c 0.0300 Ni 28058.60c 8.0641 50 60 Cr 24050.60c 0.7103 Ni 28060.60c 3.2127 52 61 Cr 24052.60c 14.2291 Ni 28061.60c 0.1420 53 62 Cr 24053.60c 1.6443 Ni 28062.60c 0.4596 54 64 Cr 24054.60c 0.4162 Ni 28064.60c 0.1216 55 Mn 25055.50c 2.0000 Mo-nat 42000.50c 2.5000 Density = 7.98 g/cm3 Source: ASM International (1987), p. 931, and ASME 1998, Section II, SA-240, Table 1.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 28 of 37 Table 5-12. Material Specifications for Neutronit A978 with 1.62 wt% Boron Element/Isotope ZAID Wt% Element/Isotope ZAID Wt% 10 57 B 5010.50c 0.2986 Fe 26057.60c 1.3928 11 58 B 5011.56c 1.3214 Fe 26058.60c 0.1872 59 C-nat 6000.50c 0.0400 Co 27059.50c 0.2000 50 58 Cr 24050.60c 0.7730 Ni 28058.60c 8.7361 52 60 Cr 24052.60c 15.4846 Ni 28060.60c 3.4805 53 61 Cr 24053.60c 1.7894 Ni 28061.60c 0.1539 54 62 Cr 24054.60c 0.4529 Ni 28062.60c 0.4979 54 64 Fe 26054.60c 3.6411 Ni 28064.60c 0.1317 56 Fe 26056.60c 59.2189 Mo-nat 42000.50c 2.2000 Density = 7.76 g/cm3 Source: MO0109RIB00049.001. Table 5-13. Material Specifications for Al 6061 Element/Isotope ZAID Wt% Element/Isotope ZAID Wt% Si-nat 14000.50c 0.6000 Mg-nat 12000.50c 1.0000 54 50 Fe 26054.60c 0.0396 Cr 24050.60c 0.0081 56 52 Fe 26056.60c 0.6433 Cr 24052.60c 0.1632 57 53 Fe 26057.60c 0.0151 Cr 24053.60c 0.0189 58 54 Fe 26058.60c 0.0020 Cr 24054.60c 0.0048 63 Cu 29063.60c 0.1884 Ti-nat 22000.50c 0.1500 65 27 Cu 29065.60c 0.0866 Al 13027.50c 96.9300 55 Mn 25055.50c 0.1500 Density = 2.7065 g/cm3 Source: MO9906RIB00048.000. 27 NOTE: Zn cross-section data unavailable, therefore it was substituted as Al (See assumption 3.7). Table 5-14. Material Specifications for Grade 70 A516 Carbon Steel Element/Isotope ZAID Wt%a Element/Isotope ZAID Wt%a 54 C-nat 6000.50c 0.2700 Fe 26054.60c 5.5558 55 56 Mn 25055.50c 1.0450 Fe 26056.60c 90.3584 31 57 P 15031.50c 0.0350 Fe 26057.60c 2.1252 32 58 S 16032.50c 0.0350 Fe 26058.60c 0.2856 Si-nat 14000.50c 0.2900 Densityb = 7.850 g/cm3 a Sources: ASTM A 516/A 516M-01 (2001), Table 1. b ASTM A 20/A20M-99a (1999), p. 9.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 29 of 37 Table 5-15. Material Specifications for Zircaloy-2 Element/Isotope ZAID Wt%a Element/Isotope ZAID Wt%a 16 58 O 6000.50c 0.1250 Fe 26058.60c 0.0004 50 58 Cr 24050.60c 0.0042 Ni 28058.60c 0.0370 52 60 Cr 24052.60c 0.0837 Ni 28060.60c 0.0147 53 61 Cr 24053.60c 0.0097 Ni 28061.60c 0.0007 54 62 Cr 24054.60c 0.0024 Ni 28062.60c 0.0021 54 64 Fe 26054.60c 0.0076 Ni 28064.60c 0.0006 56 Fe 26056.60c 0.1241 Sn-nat 50000.35c 1.4500 57 Fe 26057.60c 0.0029 Zr-nat 40000.60c 98.1350 Density = 6.55 g/cm3 Source: MO9906RIB00048.000. Table 5-16. Material Specifications for Zircaloy-4 Element/Isotope ZAID Wt% Element/Isotope ZAID Wt% 50 57 Cr 24050.60c 0.0042 Fe 26057.60c 0.0045 52 58 Cr 24052.60c 0.0837 Fe 26058.60c 0.0006 53 16 Cr 24053.60c 0.0097 O 8016.50c 0.1250 54 Cr 24054.60c 0.0024 Zr-nat 40000.60c 98.1150 54 Fe 26054.60c 0.0119 Sn-nat 50000.35c 1.4500 56 Fe 26056.60c 0.1930 Density = 6.56 g/cm3 Source: MO9906RIB00048.000.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 30 of 37
- 6. RESULTS The Limerick Generating Station Unit 1 radiochemical assay comparison results are presented in this section. The criticality calculations were performed using two different sets of compositions for the fuel material while keeping unchanged all the other material compositions and geometry parameters for the waste package disposal configuration. One set consists of the concentrations for 32 of the isotopes measured in eight samples, and the other set consists of the concentrations for the same isotopes calculated using the SAS2H sequence of SCALE 4.4a. The results presented provide a comparison of the calculated isotopic concentrations with the measured isotopic concentrations on a percent difference basis. The difference between the measured and the calculated value was divided by the measured value to determine the accuracy of the SAS2H calculation. A positive percent difference represents an over prediction by the code, while a negative percent difference represents an under prediction by the code. Two additional calculations were performed, for the two fresh fuel compositions (and using the same geometry).
In order to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on system reactivity, MCNP calculations were performed to calculate the multiplication factor (k) that results from using the different sets of isotopic concentrations and provide a comparison in terms of k. The results represent the average combined collision, absorption, and track-length estimator from the MCNP calculations. The standard deviation () represents the standard deviation of k about the average combined collision, absorption, and track-length estimate due to the Monte Carlo calculation statistics. The SAS2H and MCNP input and output files used in this evaluation are contained on an attachment compact disc to this calculation file as listed in Attachment II. The outputs are considered reasonable compared to inputs. The results are suitable for their intended use.
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 31 of 37 6.1 SAMPLE RESULTS FROM ASSEMBLY YJ1433 Table 6-1 presents the SAS2H calculated isotopic concentrations. Table 6-1. SAS2H Calculated Isotopic Concentrations for Assembly YJ1433 Samples Sample D8-3D2B D8-4G3 D9-1D2 D9 2D2 D9 4D4 D9-4G1E1 H5-3A1C H5-3A1G Node 15 23 6 11 18 22 15 16 Isotope (g/g238U) 234 U 1.49E-04 1.97E-04 1.62E-04 1.56E-04 1.50E-04 1.71E-04 1.73E-04 1.72E-04 235 U 2.63E-03 6.62E-03 2.13E-03 2.38E-03 1.68E-03 3.09E-03 4.09E-03 4.20E-03 236 U 5.57E-03 5.15E-03 6.11E-03 6.17E-03 6.18E-03 6.17E-03 6.15E-03 6.16E-03 238 Pu 4.45E-04 1.87E-04 4.27E-04 5.32E-04 4.61E-04 3.78E-04 4.62E-04 4.68E-04 239 Pu 4.85E-03 4.15E-03 4.32E-03 4.95E-03 4.17E-03 4.29E-03 5.31E-03 5.40E-03 240 Pu 3.54E-03 2.63E-03 3.33E-03 3.69E-03 3.52E-03 3.33E-03 3.58E-03 3.60E-03 241 Pu 1.26E-03 8.58E-04 1.13E-03 1.32E-03 1.11E-03 1.08E-03 1.35E-03 1.36E-03 242 Pu 1.50E-03 6.38E-04 1.61E-03 1.71E-03 1.70E-03 1.28E-03 1.33E-03 1.32E-03 143 Nd 1.04E-03 9.03E-04 1.00E-03 1.11E-03 9.79E-04 1.01E-03 1.14E-03 1.15E-03 145 Nd 1.12E-03 8.42E-04 1.20E-03 1.24E-03 1.22E-03 1.11E-03 1.14E-03 1.13E-03 146 Nd 1.40E-03 8.80E-04 1.50E-03 1.60E-03 1.60E-03 1.35E-03 1.38E-03 1.38E-03 148 Nd 6.83E-04 4.55E-04 7.27E-04 7.70E-04 7.59E-04 6.58E-04 6.78E-04 6.76E-04 150 Nd 3.48E-04 2.18E-04 3.66E-04 3.93E-04 3.85E-04 3.27E-04 3.40E-04 3.40E-04 134 Cs 6.10E-05 2.92E-05 6.31E-05 7.20E-05 7.13E-05 5.77E-05 5.86E-05 5.88E-05 137 Cs 2.06E-03 1.37E-03 2.18E-03 2.32E-03 2.29E-03 1.98E-03 2.04E-03 2.03E-03 151 Eu 5.28E-07 3.94E-07 4.79E-07 5.72E-07 4.86E-07 4.73E-07 5.89E-07 6.00E-07 153 Eu 2.37E-04 1.42E-04 2.51E-04 2.65E-04 2.58E-04 2.20E-04 2.31E-04 2.30E-04 155 Eu 6.61E-06 3.64E-06 6.84E-06 7.49E-06 7.14E-06 6.03E-06 6.45E-06 6.45E-06 147 Sm 2.78E-04 2.67E-04 3.00E-04 2.91E-04 2.85E-04 2.87E-04 2.89E-04 2.87E-04 149 Sm 2.46E-06 2.18E-06 2.24E-06 2.63E-06 2.40E-06 2.60E-06 2.69E-06 2.75E-06 150 Sm 5.26E-04 3.39E-04 5.45E-04 5.91E-04 5.71E-04 5.01E-04 5.26E-04 5.26E-04 151 Sm 1.57E-05 1.16E-05 1.43E-05 1.70E-05 1.46E-05 1.42E-05 1.74E-05 1.77E-05 152 Sm 2.23E-04 1.63E-04 2.39E-04 2.44E-04 2.46E-04 2.18E-04 2.19E-04 2.18E-04 155 Gd 5.70E-06 3.14E-06 5.88E-06 6.45E-06 6.13E-06 5.18E-06 5.58E-06 5.58E-06 241 Am 3.50E-04 2.44E-04 3.09E-04 3.64E-04 2.95E-04 2.90E-04 3.79E-04 3.84E-04 241m Am 1.26E-06 9.37E-07 9.90E-07 1.27E-06 8.59E-07 8.99E-07 1.52E-06 1.55E-06 243 Am 4.25E-04 1.25E-04 4.43E-04 5.12E-04 4.74E-04 3.29E-04 3.78E-04 3.77E-04 237 Np 7.62E-04 4.56E-04 7.70E-04 8.74E-04 7.86E-04 7.07E-04 8.13E-04 8.18E-04 95 Mo 1.32E-03 9.59E-04 1.42E-03 1.47E-03 1.46E-03 1.31E-03 1.33E-03 1.33E-03 99 Tc 1.38E-03 1.00E-03 1.48E-03 1.53E-03 1.51E-03 1.36E-03 1.39E-03 1.38E-03 101 Ru 1.42E-03 9.58E-04 1.52E-03 1.59E-03 1.57E-03 1.37E-03 1.41E-03 1.40E-03 103 Rh 7.21E-04 5.42E-04 7.35E-04 7.73E-04 7.27E-04 6.81E-04 7.29E-04 7.28E-04 109 Ag 1.82E-04 1.02E-04 1.87E-04 2.02E-04 1.94E-04 1.60E-04 1.73E-04 1.72E-04 242 Cm 4.71E-08 2.83E-08 4.62E-08 5.10E-08 4.45E-08 4.32E-08 4.83E-08 4.90E-08 243 Cm 9.82E-07 3.76E-07 8.87E-07 1.11E-06 9.33E-07 8.19E-07 9.91E-07 1.01E-06 244 Cm 1.99E-04 3.22E-05 1.96E-04 2.63E-04 2.44E-04 1.40E-04 1.68E-04 1.68E-04 245 Cm 8.21E-06 1.02E-06 6.73E-06 1.11E-05 8.77E-06 5.17E-06 7.54E-06 7.74E-06
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Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 32 of 37 Table 6-2 presents the results in terms of percent difference for the calculated isotopic concentrations compared with the measured results (Table 5-2). Table 6-2. Assembly YJ1433 Sample Percent Differences Sample D8-3D2B D8-4G3 D9-1D2 D9 2D2 D9 4D4 D9-4G1E1 H5-3A1C H5-3A1G Node 15 23 6 11 18 22 15 16 Isotope (Calculated Concentration - Measured Concentration) x 100/Measured Concentration 234 U -7.42 1.29 -2.46 -4.58 -9.08 -7.70 -4.15 -4.18 235 U -38.29 -24.06 25.07 8.82 -37.71 -28.20 -14.78 -13.97 236 U -2.49 -0.48 1.92 2.65 2.09 2.60 -0.94 -0.87 238 Pu 1.43 -26.51 6.86 4.77 -15.10 -14.99 -8.79 -7.94 239 Pu -11.15 -24.83 9.69 3.68 -21.40 -21.08 -13.45 -12.61 240 Pu -4.36 -9.15 6.73 5.41 -2.40 -0.61 -4.82 -4.45 241 Pu -8.47 -24.75 6.69 1.08 -22.65 -22.43 -10.29 -9.08 242 Pu 26.02 0.81 4.32 5.53 8.91 3.71 14.98 16.01 143 Nd 1.81 -1.76 17.36 12.49 -5.83 -2.60 2.67 3.34 145 Nd 11.07 5.39 6.10 5.92 5.20 5.05 6.22 5.97 146 Nd 13.67 3.25 4.98 5.48 6.52 5.14 4.42 5.05 148 Nd 11.06 2.56 4.45 4.56 4.14 3.32 4.09 4.23 150 Nd 12.60 0.85 7.05 6.68 5.48 3.69 4.77 5.26 134 Cs 6.17 -12.86 7.08 0.38 -1.25 1.99 -6.14 -5.27 137 Cs 9.75 1.61 9.55 6.62 9.30 11.13 4.24 4.60 151 Eu 10.62 -8.73 40.03 41.64 10.99 4.64 13.44 14.05 153 Eu 22.21 1.03 20.23 20.06 19.85 9.35 17.11 16.76 155 Eu -16.92 -32.15 -14.57 -16.91 -18.27 -24.57 -20.45 -20.34 147 Sm -3.55 3.42 1.86 -1.82 -1.71 -4.11 -1.72 -2.58 149 Sm -8.41 -24.92 33.99 11.94 -14.44 -15.57 -8.47 -6.67 150 Sm 9.32 1.64 9.52 7.22 5.12 2.19 3.68 4.34 151 Sm 16.04 -5.42 41.50 37.36 7.96 4.97 12.84 15.72 152 Sm 46.44 35.72 33.68 38.76 48.12 38.11 46.94 47.24 155 Gd -52.11 -67.61 -24.62 -26.59 -33.28 -33.56 -35.12 -30.99 241 Am -10.07 -23.16 9.22 11.64 -22.54 -21.54 -7.02 -7.23 241m Am 6.60 -19.25 48.26 55.83 -13.22 -15.18 7.19 6.32 243 Am 40.41 -3.71 22.30 30.55 13.12 9.33 27.68 26.81 237 Np -3.74 -15.62 0.35 -2.91 -11.25 -13.96 -5.47 -5.46 95 Mo 3.72 -3.51 -1.04 3.47 6.37 2.97 3.81 3.58 99 Tc 8.87 0.17 6.35 12.66 8.18 17.11 13.64 17.20 101 Ru 11.07 0.23 1.11 5.44 7.09 2.97 6.51 3.13 103 Rh -3.23 -10.73 1.78 1.88 -9.12 -8.15 -3.34 -7.10 109 Ag 13.10 -7.30 28.02 31.71 24.95 32.60 28.78 14.01 242 Cm -4.01 -24.12 23.20 35.45 -15.68 -22.24 0.34 14.86 243 Cm -14.65 -41.90 4.53 -1.57 -31.90 -32.31 -20.11 -22.32 244 Cm 27.45 -30.14 8.31 19.07 0.45 -7.46 3.40 3.16 245 Cm -35.88 -69.04 -29.19 -27.70 -53.86 -54.28 -50.70 -49.38
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 33 of 37 6.2 MCNP CALCULATION RESULTS The results of MCNP runs using standard fuel rods with the fresh fuel compositions from Table 5-7 are presented in Table 6-3. Table 6-3. MCNP Results for Fresh Nuclear Fuel Pin Type Filling the 235 Waste Package U (Wt%) keff AENCF D8 3.60 (with 5.0 wt% Gd2O3) 0.17581 0.00014 0.827 D9 3.95 0.91651 0.00068 0.163 Table 6-4 presents the MCNP results for each of the spent nuclear fuel samples. Table 6-4. MCNP Results for Spent Nuclear Fuel Samples Delta (Operating History - RCAa Operating Historyb c Sample # Sample ID RCA) keff AENCF keff AENCF keff d 1 D8-3D2B 0.55913 0.00052 0.304 0.51709 0.00045 0.330 -0.04204 0.00069 2 D8-4G3 0.63515 0.00052 0.259 0.57645 0.00049 0.279 -0.05870 0.00071 3 D9-1D2 0.46612 0.00044 0.369 0.48665 0.00041 0.351 0.02053 0.00060 4 D9-2D2 0.50712 0.00047 0.339 0.51048 0.00043 0.335 0.00336 0.00064 5 D9-4D4 0.53262 0.00049 0.324 0.46549 0.00041 0.361 -0.06713 0.00064 6 D9-4G1E1 0.56781 0.00048 0.299 0.50364 0.00048 0.331 -0.06417 0.00068 7 H5-3A1C 0.58889 0.00047 0.290 0.5552 0.00042 0.305 -0.03369 0.00063 8 H5-3A1G 0.59134 0.00051 0.291 0.55936 0.00050 0.304 -0.03198 0.00071 a NOTES: Results based on measured compositions of spent nuclear fuel samples. b Results based on SAS2H calculated compositions using reactor operating history information adjusted to yield the sample measured burnup. c Values reported were calculated from MCNP reported values which lists results out to five significant digits, therefore computing delta values from table values will introduce some roundoff error. d Delta value equals the square root of the sum of the squared values for RCAs and operating history.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 34 of 37
- 7. REFERENCES 7.1 DOCUMENTS CITED Audi, G. and Wapstra, A.H. 1995. Atomic Mass Adjustment, Mass List for Analysis. [Upton, New York: Brookhaven National Laboratory, National Nuclear Data Center]. TIC: 242718.
Briesmeister, J.F., ed. 1997. MCNP-A General Monte Carlo N-Particle Transport Code. LA-12625-M, Version 4B. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: MOL.19980624.0328. BSC (Bechtel SAIC Company) 2001. 44 BWR Waste Package Loading Curve Evaluation. CAL-UDC-NU-000005 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20011114.0132. BSC (Bechtel SAIC Company) 2002. Technical Work Plan for: Risk and Criticality Department. TWP-EBS-MD-000014 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20021209.0011. CRWMS M&O 1998. Software Qualification Report for MCNP Version 4B2, A General Monte Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 30033-2003, Rev. 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19980622.0637. CRWMS M&O 1999. Summary Report of Commercial Reactor Criticality Data for LaSalle Unit
- 1. B00000000-01717-5705-00138 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC:
MOL.19990923.0233. CRWMS M&O 2000a. Users Manual for SCALE-4.4A. 10129-UM-4.4A-00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20001130.0136. CRWMS M&O 2000b. Validation Test Report (VTR) for SCALE-4.4A. 10129-VTR-4.4A-00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20001130.0139. DOE (U.S. Department of Energy) 1992. Characteristics of Potential Repository Wastes. DOE/RW-0184-R1. Four volumes. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: HQO.19920827.0001; HQO.19920827.0002; HQO.19920827.0003; HQO.19920827.0004. DOE (U.S. Department of Energy) 1996. Spent Nuclear Fuel Discharges from U.S. Reactors 1994. SR/CNEAF/96-01. Washington, D.C.: U.S. Department of Energy. TIC: 232923. DOE (U.S. Department of Energy) 2003. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 13. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20030422.0003.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 35 of 37 Hagrman, D.L.; Reymann, G.A.; and Mason, R.E., eds. 1981. MATPRO - Version 11 (Revision 2), A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior. NUREG/CR-0497, Rev. 2. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 209823. Keenan, J.H.; Keyes, F.G.; Hill, P.G.; and Moore, J.G. 1969. Steam Tables, Thermodynamic Properties of Water Including Vapor, Liquid, and Solid Phases (English Units). New York, New York: John Wiley & Sons. TIC: 246766. Martinez, C. 2001. "RE: YMP RCA Analysis." E-mail from C. Martinez to J. Scaglione, August 2, 2001. ACC: MOL.20030731.0363. Nuclear Engineering International. 1998. World Nuclear Industry Handbook. Wilmington, Kent, England: Wilmington Business Publishing. TIC: 237121. Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. Nuclides and Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: General Electric Company and KAPL, Inc. TIC: 233705. Punatar, M.K. 2001. Summary Report of Commercial Reactor Criticality Data for Grand Gulf Unit 1. TDR-UDC-NU-000002 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20011008.0008. Reager, R. 2003. BWR Spent Fuel Isotopic Characterization, Final Report. NEDO-33094, Rev.
- 0. Sunol, California: GE Nuclear Energy, Vallecitos Nuclear Center. ACC:
MOL.20030528.0184. Scaglione, J.M. 2003. "Transmittal of Assembly and Burnup Information for Limerick Spent Nuclear Fuel Assemblies." Interoffice memorandum from J.M. Scaglione (BSC) to D.A. Thomas, July 8, 2003, 0707037983, with attachment. ACC: MOL.20030708.0230. Todreas, N.E. and Kazimi, M.S. 1990. Nuclear Systems I, Thermal Hydraulic Fundamentals. New York, New York: Hemisphere Publishing. TIC: 226511. Walpole, R.E.; Myers, R.H.; and Myers, S.L. 1998. Probability and Statistics for Engineers and Scientists. 6th Edition. Upper Saddle River, New Jersey: Prentice Hall. TIC: 242020. 7.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES AP-3.12Q, Rev. 2, ICN 0. Design Calculations and Analyses. Washington, D.C.: U.S. Department of Energy. ACC: DOC.20030403.0003. ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. Metals Park, Ohio: ASM International. TIC: 209807.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Page 36 of 37 ASME (American Society of Mechanical Engineers) 1998. 1998 ASME Boiler and Pressure Vessel Code. 1998 Edition with 1999 and 2000 Addenda. New York, New York: American Society of Mechanical Engineers. TIC: 247429. ASTM A 20/A20M-99a. 1999. Standard Specification for General Requirements for Steel Plates for Pressure Vessels. West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 247403. ASTM A 516/A 516M-01. 2001. Standard Specification for Pressure Vessel Plates, Carbon Steel, for Moderate- and Lower-Temperature Service. West Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 253997. 7.3 SOURCE DATA LISTED BY DATA TRACKING NUMBER MO0003RIB00071.000. Physical and Chemical Characteristics of Alloy 22. Submittal date: 03/13/2000. MO0109RIB00049.001. Waste Package Material Properties: Neutron Absorbing Materials. Submittal date: 09/17/2001. MO9906RIB00048.000. Waste Package Material Properties: Waste Form Materials. Submittal date: 6/9/1999. 7.4 SOFTWARE CODES CRWMS M&O 1998. Software Code: MCNP. 4B2LV. HP. 30033 V4B2LV. CRWMS M&O 2000. Software Code: SCALE. V4.4A. HP. 10129-4.4A-00.
Engineered Systems Project Calculation
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV OOA Page 37 of37
- 8. ATTACHMENTS Table 8-1 presents the attachment specifications for this calculation file. The contents of Attachment II are provided electronically <>p. an attachment CD to this calculation file. A listing of the contents of the CD is provided in hard-copy fonn for Attachment II in the attachments to this calculation file.
Table 8.1. Attachment Usting Attachment # # of PaQes _~ Description I .2*3l1ta* ~ 44-BWR Waste Package Configurations for Site Recommendation Sketch Excel spreadsheet SAS2H input files. SAS2H outputs. MCNP inputs. and II ftz.J~'3 MeN? outputs fir N/A Compact Disc attachment containing information listed in Attachment II
v- ZOOOOO-3W-}On-9MO 7 6 5 4 3 2
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o o TIHE SHEET INDEX ~ N o DRAWING NUMBER REV SHEET NUMBER o OUTER BARR IER o o o D DWG-UDC-ME-000002 A SHEET I OF D DWG-UDC-ME-000002 DWG-UDC-ME-000002 A A SHE ET 2 OF SHE ET 3 OF U
~
§ INNER ShELL J
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u FUEL BASKET ASSEMBLY c c INNER SHELL LID OUTER BARRIER FLAT CLOSURE LID B OUTER BARRIER EXTENDED LID o B o W)MINAL LENGTH NOMINAL DIAMETER LOADED WEIGIIT 5165 MM - 1674 MM 43000 KG 203 3 IN 65. 9 IN 95000 LB l'lSI',N II/PUIS
....C(
4-; THIS ORAWING IS PRELI~INARY AND NOT I NTENDED FOR CONSTRUCT ION, PROCURE~ENT SfE OO(..".. r tIT IIIPUT 0 OR FABRICATION. R(f(R(H([ SYSI[N IOtRSI NorES: A I. ALL DIMENSIONS AIlE IN MILL IMETERS ..... -"------- -'~'*U.S. DEPARTMENT
,:,~.,.tl\ !Yucca Mountain Site OF ENERGY A Characterization P.ro*ect (1) 00 EXCEPT BRACKETED AilE IN INCIIES D. ;"(KEllll [
r--*UHCH'J'HL (lfll,OIl Ihd,('ocl"t tljQ~tt M'l'C~m~nt ~e'trn C'<l p,n"" SAl ~ MArIAGlM[NT & OPf.HATlnG orlIRA.IOR p.. D. tl(K(tlll [
- 2. ALL MASSES ARE ROUNDED uP 10 (Hiw: REPOSI lORY DESIGN THE NEAREST THOUSAND T. : ou" I iT
-r.'ll~l~ WASTE PACKAGE PROJECT E
- 3. DO NOT SCALE FROM DRAWING ,,-t---
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SHEET I OF 3 .§
- 4. OUHR BARRIER AND INNER SIIELL -A U
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[ 65 9 J BASKET B-SIDEGUIDE o o
¢ 1594 REF OUTER BARRIER OD BASKET B-STIFfENER >
[62 8] FUEL BASKET G*PLATE (THERMAL SHUNT) ~ N o
¢305 REF OUTER BARRIER LID LIFTING FEATURE OD FUEL BASKET E-PLATE o
[ 12 . OJ o o o D BASKET STIFFENER D A FUEL BASKET B-PLATE / ; ~
.J BASKET CORNERGUIDE U
§ FUEL BASKET D-PLATE J
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u o ¢ 1554 [6 I 2 ] REF OUTER BARR IER ID FUEL BASKET C-PLATE ¢ 1554 REF INNER SHELL OD ( 6 I 21
¢ 1454 REF INNER SHELL ID
[ 57.21 c 1--------------------1-¢1571 [61 9 ) REF OUTER BARRIER EXTENDED LID FUEL BASKET F-PLATE THERMAL (SHUNT) C 11--------------------II-¢1557 REF OUTER BARRIER CLOSURE LID [ 6 I 3J SEE DETAIL A SEE SHEET 3 SEE DETAIL C SEE SHEET 3 o
.FUEL BASKET TUBl 44 PLACES lB 8 8 4775 REF INNER SHELL
[ 188 ] 5165 REF NOMINAL LENGTH 4585 REF CAVITY LENGTH [ 203 3 J [180 5 J 4575 REF BASKET LENGTH 155 TVP FUEL BASKET C*PLATE [180 I J [6. 1] FUEL BASKET A*PLATE SECTION B-B SCALE 1:3 THIS DRAWING IS PRELIMINARY AND NOT l'lSIG'l ItlPUIS - M I 4 -; o I NTENDED fOR CONSTRUCT ION, PROCUREMENT S[( DOCUM£NT It/PuT OR fABRICATION. R[f(A(N(( SY$I04 ID1RSI N I A NOTE S: .."....,." U.S. DEPARTMENT OF ENERGY A I. ALL DIMENSIONS ARE IN MILLIMETERS D. f.lCK[ III I ( O"'\~,.I\ Yucca Mountain Site Characterization Pro'eet
<l) 0/}
EXCEPT (JRACKETED ARE IN INCHES OiiG"OiOfl !~-- '~'Kf~~ ~1;f~~(N~~~r~Tooll'Ohlr~~jTe, ~g)to~~~kltVo~ o;l Q., D. rICK[ tlll ( 1....."".. - REPOSIIORY DESIGN
- 2. ALL MASSES ARE ROUNDED UP TO
~
(~f(i: 0 SECTION A-A THE NEAREST THOUSAND T. iCHMITT --f;1I[..k WASTE PACKAGE PROJECT SEE DETAIL B
~~:,~.~;",." *.-+---1 44-BWR WASTE PACKAGE SEE SHEET 3 SCALE 1:10 3. DO NOT SCALE FROM DRAWING SHEET 2 OF 3
- 4. OUTER BARRIER AND INNER SHEI L -A IIAVE RAD IAL GAP OF 0 TO 4 :'IM REF 44 BWR.BASElINE 8 7 6 5 4 :'j'\ '-:.:\
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u 125 r [4 9] REF
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REF BASKET TO INNER LID GAP 25 REF [ I 0] 30 REF CLOSURE LID TO OUTER LID GAP DETAIL C [ I 2] SCALE 1:3 L 100 REF [3.9] c FL-...,-~~<-----~~~ CLOSURE LID WELD f C [ 4] 10 REF CLOSURE LID 340 [ 4J [ I3 30 REF INNER LID TO CLOSURE LID GAP [I 2] OTY COMPOIllENT MATERIAL MASS ROD Roo BASKET B-SIDEGUIDE SA -516 K02700 32 REF INNER SHELL LID BASKET B-SfIFFENER SA-516 K02700 64 D.n 345 REF R:F REF BASKET C0RNE~GUIDE SA-SI6 K02700 32 [136J 13.9J 1_ 0] BASKET SIIFFENER SA-516 K02700 64 FUEL BASKEI A-PLATE NEU1RONIT A 978 4 fUEL BASKEl B-pLATE NE UIRON I I A 978 4 225 REF IUEL BASKET (-PLATE NEUIRONIT A 918 16 [8 9 J B FUEL BASKE: D-pLAIE NEUTIlONIT A 978 16 B FUEL BASKE' E-PLATE NEUIIlONIT A 918 16 I UEL BASKE' F-I'LATE SB*209 A96061 T4 8 FUEL BASKE G-PLAIE 56-209 A96061 T4 8 FUEL BASKLT rUSE SA-516 K02700 44 INNER :,IIELL INNER SIlILL LID SA-240 S31600 SA-210 S31600 I I 20 REr - [ 8J
-1 20 REF-- IUNER SII ELL LID I 1FT I NG I EATURE SA*210 S31600 I
[8 J 2] REF OUTER Sf,RRIER SA*575 N06022 I DETAIL B 20 OUTER CARR IER EX TEN['ED LID SA-57S N06022 I SCALE 1:2 [8 J 60 REF - - I
--- ~~ {)UIER BARRIER LID LII liNG IEAIURE
':'UIlR l\AIlIlIER FLAT (L('SURE LID SA-SIS N06022 SA-SIS N06022 2
I
.l2.LO..l ) (d
[2 4 J ,- ~O REF 21000 KG I EFt E!'.C='=T~IV'=E-=D"";"T;::£'--- [2 0 J WASTE PACKAGE ASSEMBLY 60000 LB I 43000 KG I l'lSIl;N IIlPUIS
'NP ASSEMBL Y WI 1H SNF THIS DRAWING IS PRELIMINARY AND NOT 95000 LB I I NTENDED FOR CONSTRUCT ION. PROCUREMENT OR FABRICATION.
SEE DOCtJ~!tn RU[R(NC[ srSl(fll IDIR$I IIJPUT NOH S A I ALL DIMENSIONS ARE IN MILLIMETERS U.S. DEPARTMENT OF ENERGY A EXCEPT BI(ACKETED AIlE IN INCIIES O. M(K[ Ilil E
~,~" Yucca Mountain Site Characterization Pro 'ect ii';C;,*';'OI'---
O.
~....J.' nXfl-E!!:L ~'A"I~Go[~~a~T(><)& t 'O~lRl\el ~Gn'\~o~1Ikll:.~,'oR
- 2. ALL MASSES ARE ROUNDED UP TO M(K[ HZ I (
(';;<<.'0---- - B - DETAIL A REPOSI lORY UESIGN THE NEAREST THOUSAND T. sell'" I TT
".,'.(01'011--- 1
,~~ WASTE PACKAGE PROJECT SCALE 1:2 3. DO NOT SCALE FROM DRAWING tUA 44-BWR WASTE PACKAGE
- s. REtltlE TT 'S,..o,'l~\ SHEET 3 OF 3
- 4. OUTER OARRIER AND INNER SHELL <>{; .. \",1o, HAVE RADIAL GAP OF 0 TO 4 :.1101 REF fvC., D. crT"",IEii,:,'_ _ J-,;,;;;+---,-,;d<a7.7,-;i,;.r.-,,..,::-:=-,T.ii-;;-i""'--'::"::'::":":";'---'-'--jA J. (LOUD 44-0WR_BASEL IllE 8 flU U R"'LIIASrLIN! IlhllNG 7 6 5 4 2
Engineered Systems Project Calculation Attachment
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Attachment II, Page II-1 of II-2 ATTACHMENT II This attachment contains a listing and description of the zip file contained on Attachment III, the CD of this calculation. The CD was written using the Hewlett Packard (HP) CD-Writer Plus model 7200e external CD-rewritable drive for personal computers, and the zip archive was created using WINZIP 8.1. The zip file attributes on the CD are as follows: Archive Filename File Size (bytes) File Date File Time att.zip 5,056,040 07-28-2003 4:17p There are 78 total files contained in the root directory. Upon file extraction, the file naming system corresponds as follows for the SAS2H cases, and as listed in Table II-1 for the MCNP cases.
- N*.inp files are the SAS2H input files.
- N*.msgs files contain the standard run-time messages associated with the SAS2H calculations (these are generated by SAS2H).
- ft72f001.N* files are temporary ASCII files generated by SAS2H, which must be retained, that contain the isotopic concentrations as a function of time (the actual SAS2H output file contains a large amount of information that is not needed for this calculation, therefore it is discarded, but the temporary files SAS2H creates are retained).
- act_N*.mass files contain the extracted actinide isotopes from the ft72f001.N* files and provides them in units of grams.
- fp_N*.mass files contain the extracted fission product isotopes from the ft72f001.N* files and provides them in units of grams.
Table II-1. Sample and MCNP Filename Identification File Namea Sample ID Comments 1m D8-3D2B 2m D8-4G3 3m D9-1D2 Cases for samples from assembly YJ1433 4m D9-2D2 radiochemical assay 5m D9-4D4 measured isotopic 6m D9-4G1E1 concentrations. 7m H5-3A1C 8m H5-3A1G 1c D8-3D2B 2c D8-4G3 Cases for samples from 3c D9-1D2 assembly YJ1433 using 4c D9-2D2 operating history information 5c D9-4D4 adjusted to sample measured 6c D9-4G1E1 burnup in SAS2H to generate 7c H5-3A1C isotopic concentrations. 8c H5-3A1G 1f Representative for pin D8 Cases with fresh fuel in the Representative for pins same geometric configuration 2f D9 and H5 as the rest of the cases. a NOTE: Input files have an i at the end of the file name, and output files have an io at the end of the file name.
Engineered Systems Project Calculation Attachment
Title:
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations Document Identifier: CAL-DSU-NU-000002 REV 00A Attachment II, Page II-1 of II-2 LGS1RCA.xls is an Excel spreadsheet that contains design parameters and operation information (limited to cycles 5, 6, and 7) for Limerick Unit 1 BWR core and YJ1433 fuel assembly.}}