PSR-0210 SCOPE.
SCOPE, Shipping Cask Optimization and Parametric Evaluation
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| SCOPE | PSR-0210/01 | Arrived | 30-JAN-2002 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| PSR-0210/01 | IBM 3033 |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
Given the neutron and gamma-ray shielding requirements as input, SCOPE may be used as a conceptual design tool for the evaluation of various casks designed to carry square fuel assemblies, circular canisters of nuclear waste material, or circular canisters containing "intact" spent-fuel assemblies. It may be used to evaluate a specific design or to search for the maximum number of full assemblies (or canisters) that might be shipped in a given type of cask. In the "search" mode, SCOPE will use built-in packing arrangements and the tabulated shielding requirements input by the user to "design" a cask carrying one fuel assembly (or canister); it will then continue to increment the number of assemblies (or canisters) until one or more of the design limits can no longer be met. In each case (N = 1,2,3...), SCOPE will calculate the steady-state temperature distribution throughout the cask and perform a complete 1-D space/time transient thermal analysis following a postulated half-hour fire; then it will edit the characteristic dimensions of the cask (including fins, if required), the total weight of the loaded case, the steady-state temperature distribution at selected points, and the maximum transient temperature in key components. With SCOPE, the effects of various design changes may be evaluated quickly and inexpensively.
Given the neutron and gamma-ray shielding requirements as input, SCOPE may be used as a conceptual design tool for the evaluation of various casks designed to carry square fuel assemblies, circular canisters of nuclear waste material, or circular canisters containing "intact" spent-fuel assemblies. It may be used to evaluate a specific design or to search for the maximum number of full assemblies (or canisters) that might be shipped in a given type of cask. In the "search" mode, SCOPE will use built-in packing arrangements and the tabulated shielding requirements input by the user to "design" a cask carrying one fuel assembly (or canister); it will then continue to increment the number of assemblies (or canisters) until one or more of the design limits can no longer be met. In each case (N = 1,2,3...), SCOPE will calculate the steady-state temperature distribution throughout the cask and perform a complete 1-D space/time transient thermal analysis following a postulated half-hour fire; then it will edit the characteristic dimensions of the cask (including fins, if required), the total weight of the loaded case, the steady-state temperature distribution at selected points, and the maximum transient temperature in key components. With SCOPE, the effects of various design changes may be evaluated quickly and inexpensively.
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4. METHOD OF SOLUTION
SCOPE assumes that the user has already made an independent determination of the neutron and gamma-ray shielding requirements for the particular type of cask(s) under study. The amount of shielding required obviously depends on the type of spent fuel or nuclear waste material, its burnup and/or exposure, the decay time, and the number of assemblies or canisters in the cask. Source terms (and spectra) for spent PWR and BWR fuel assemblies are provided at each of 17 decay times, along with recommended neutron and gamma-ray shield thicknesses for Pb, Fe, and U-metal casks containing a number (1-25) of 1-, 2-, 3-, 4-, 7-, or 10-year-old PWR spent fuel assemblies.
SCOPE assumes that the user has already made an independent determination of the neutron and gamma-ray shielding requirements for the particular type of cask(s) under study. The amount of shielding required obviously depends on the type of spent fuel or nuclear waste material, its burnup and/or exposure, the decay time, and the number of assemblies or canisters in the cask. Source terms (and spectra) for spent PWR and BWR fuel assemblies are provided at each of 17 decay times, along with recommended neutron and gamma-ray shield thicknesses for Pb, Fe, and U-metal casks containing a number (1-25) of 1-, 2-, 3-, 4-, 7-, or 10-year-old PWR spent fuel assemblies.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
The key physical features of those casks considered by the SCOPE code include: an inner steel shell, a gamma-ray shield, an outer steel shell, a neutron shield, and an outside barrel. Depending on the amount of decay heat that must be dissipated, the cask(s) may or may not have circumferential fins. Inside the cask(s), the spent fuel assemblies (or canisters) may be separated by means of an aluminium or copper insert. It is assumed that the spent fuel is to be shipped dry (i.e., casks with forced circulation cooling systems are not considered). Lead, iron, or uranium metal may be specified as the gamma-ray shielding material, while the neutron shield is always assumed to be a common mixture of water and ethylene glycol containing approximately 1 wt% boron.
The key physical features of those casks considered by the SCOPE code include: an inner steel shell, a gamma-ray shield, an outer steel shell, a neutron shield, and an outside barrel. Depending on the amount of decay heat that must be dissipated, the cask(s) may or may not have circumferential fins. Inside the cask(s), the spent fuel assemblies (or canisters) may be separated by means of an aluminium or copper insert. It is assumed that the spent fuel is to be shipped dry (i.e., casks with forced circulation cooling systems are not considered). Lead, iron, or uranium metal may be specified as the gamma-ray shielding material, while the neutron shield is always assumed to be a common mixture of water and ethylene glycol containing approximately 1 wt% boron.
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10. REFERENCES
- SCALE: A Modular Code System for Performing Standarized Analyses
for Licensing Evaluation,
NUREG/CR-0200 (ORNL/NUREG/CSD-2), Vols I, II, and III (april 1982)
- SCALE: A Modular Code System for Performing Standarized Analyses
for Licensing Evaluation,
NUREG/CR-0200 (ORNL/NUREG/CSD-2), Vols I, II, and III (april 1982)
PSR-0210/01, included references:
- J.A. Bucholz:Scoping Design Analysis for Optimized Shipping Casks Containing
1-, 2-, 3-, 5-, 7-, or 10-Year Old PWR Spent Fuel.
ORNL/CSD/TM-149 (January 1983)
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PSR-0210/01
source program mag tapeSCOPE Source Program (Fortran-IV) SRCTPsource program mag tapeInput Processor Source Program (Fortran-IV)SRCTP
test-case data mag tapeSCOPE JCL to Compile and Save HEX Deck DATTP
test-case data mag tapeSCOPE HEX Deck for SCOPE Code DATTP
test-case data mag tapeSCOPE Input Data File for Sample Problem DATTP
test-case data mag tapeSCOPE JCL to Load HEX Deck and Process DataDATTP
test-case output mag tapeSCOPE Test Case Printed Output OUTTP
report ORNL/CSD/TM-149 (January 1983) REPPT
Keywords: radioactive waste storage, radioactivity transport, shielding.