ML20343A208
| ML20343A208 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/2020 |
| From: | Borowski J, Michel Call, Darrell Dunn, Antonio Rigato, James Smith, Jorge Solis, Jeremy Tapp, Bernie White Division of Fuel Management |
| To: | |
| SKumar - NMSS/DFM/NARAB | |
| Shared Package | |
| ML20343A188 | List: |
| References | |
| NUREG-2216 | |
| Download: ML20343A208 (12) | |
Text
APPENDIX D BENCHMARK CONSIDERATIONS FOR MIXED OXIDE RADIOACTIVE MATERIALS AND SPENT NUCLEAR FUEL D.1 Experimental Benchmarks The information and guidance in this appendix applies to both mixed oxide (MOX) radioactive materials and spent nuclear fuel (SNF) packages. This appendix does not address considerations for burnup credit for commercial MOX SNF, whether irradiated in a pressurized-water reactor or a boiling-water reactor; the considerations are for analyses that assume the MOX fuel is unirradiated. Benchmarking for any commercial MOX SNF would need to address additional considerations, such as those indicated in the discussion about MOX burnup credit in Section 6.4.7 of this SRP.
Substantial guidance on how to select an appropriate set of criticality benchmark experiments for low-enriched uranium (LEU) fissile systems is given in NUREG/CR-5661, Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, issued April 1997 (Dyer and Parks 1997), and in NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, issued March 1997 (Lichtenwalter et al. 1997). Considerably fewer benchmark experiments exist for MOX than for LEU. As a consequence, the guidance provided in NUREG/CR-5661 and NUREG/CR-6361 cannot be applied directly to the evaluation of MOX fissile systems. The benchmarks needed for the criticality analyses of MOX packages are in the thermal energy range. This condition results because, for essentially all types of MOX, the most reactive configuration is a flooded containment.
As an alternative, the 2001 edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) has 11 evaluated thermal-energy studies involving MOX fuel pins in various lattice experiments and five evaluated thermal-energy studies involving MOX liquids in tank experiments (NEA, 2001). These can be divided into 18 sets of experiments involving different fissile oxide compositions and configurations in lattices and 13 sets of experiments involving different liquid fissile nitrate compositions and configurations in tanks.
The total number of essentially different experiments is 131. Since the 2001 edition, an additional four evaluated thermal-energy studies involving MOX fuel pins and an additional four evaluated thermal-energy studies involving MOX liquids have been added to the IHECSBE (NEA, 2014) that include experiments evaluated to be acceptable to use as benchmarks. Other benchmark experiments are available throughout the world but are not as readily available. The vast majority have not been rigorously evaluated in the manner of those found in the IHECSBE and are consequently of limited use for benchmark criticality analyses for MOX packages. More evaluated MOX thermal benchmarks may be included in future editions of the IHECSBE.
The 18 sets of experiments involving fissile oxides in lattices and 13 sets of experiments involving fissile nitrate liquids in tanks from the 2001 edition of the IHECSBE have been organized and shown in Tables D-1 through D-5. The various tables are separated on two features. The first is between lattice and tank experiments, and the second is on weight percent of plutonium to total plutonium plus uranium (Pu/(Pu+U). Table D-1 has lattice experiments with Pu/(Pu+U) to 5 percent. Table D-2 has lattice experiments with Pu/(Pu+U) from 5 percent to 15 percent. Table D-3 has lattice experiments with Pu/(Pu+U) greater than 15 percent.
Table D-4 has tank experiments with Pu/(Pu+U) to 31 percent (there are no experiments with Pu/(Pu+U) less than 22 percent). Table D-5 has tank experiments with Pu/(Pu+U) greater than D-1
31 percent. Lists of meaningful, experimental characteristics are recorded for each set of experiments together with characteristics of their corresponding computational evaluations.
Experimental plutonium benchmarks should also be taken into account as part of the initial set of benchmark experiments to be considered for a MOX package application. About four times as many thermal-plutonium-tank-liquid benchmarks exist in the IHECSBE as thermal-MOX-tank-liquid benchmarks. However, fewer thermal-plutonium-lattice benchmarks exist in the IHECSBE than thermal-MOX-lattice benchmarks.
Also, there is a set of 156 configurations known as the French Haut Taux de Combustion (HTC) experiments. The descriptions of these experiments are provided in the four reports by Fernex listed in Section D3.0 and are considered commercial proprietary. Note that these experiments were set up to simulate the isotopic compositions of irradiated LEU fuel; so, the compositions will not be the same as for MOX fuel and will include other radionuclides that are not present in MOX fuel. Thus, use of the HTC experiments requires appropriate consideration of the differences between the HTC compositions and those of MOX fuel, whether irradiated and unirradiated. An evaluation of the HTC experiment data is described in NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, issued September 2008, though this evaluation was done for the purpose of using the data to benchmark burnup credit analyses for LEU SNF.
D.2 Summary of Bias and Uncertainty Evaluation There are two measures of the accuracy of an experiment and its associated calculation. The first measure is the effective bias (Eff-Bias) between calculation and benchmark experiment.
The multiplication coefficient for a fissile system is designated as keff. Designate the calculated keff for the benchmark experiment as kcalc and the benchmark experimental keff as kexp. If the calculational bias, , is defined as = kcalc - kexp, then a quantity k can be defined as follows:
(D-1) k=
0 >
or a given experimental benchmark set, kmax is chosen as the largest absolute value of the k given by Equation D-1 for all experiments in the set. The 95 percent confidence limit of kcalc is kcalc plus twice the calculated standard deviation, which is designated by 2. The Eff-Bias value is then given by the following:
Eff-Bias = kmax - 2 (D-2)
Eff-Bias, as defined here, is always less than zero. If kcalc is greater than kexp for all experiments in a set, the Eff-Bias value is just the negative of twice the calculated standard deviation.
The second measure is the total experimental uncertainty (Exp-Uncer) that was determined by the evaluator after assessing all sources of uncertainty for the experiments in a set. 1 A 13F worst-case difference between kcalc and kexp can be assigned as the difference of the total experimental uncertainty and the effective bias (Exp-Uncer - Eff-Bias) for the experimental set in question. This worst-case difference (WCD), as defined here, is always greater than zero. It represents the upper limit of the inherent uncertainties in the ability of the computer code, 1 The evaluator included sources of experimental bias or error in each kexp. This does not represent an uncertainty and so is not included in the value for total experimental uncertainty.
D-2
together with the cross-section set used, to accurately determine the keff of a critical benchmark experiment. Therefore, a bounding multiplication coefficient, ksafe, at the 95 percent confidence limit, can be chosen to be equal to 0.95 minus WCD, where an administrative margin of safety of 0.05 has been included. 2 14F Values for the variable WCD for each experimental set vary between 0.0071 to 0.0192 (0.71 percent to 1.92 percent), 0.0043 to 0.0328 (0.43 percent to 3.28 percent), 0.0023 to 0.0138 (0.23 percent to 1.38 percent), 0.0044 to 0.0180 (0.44 percent to 1.80 percent), and 0.0044 to 0.0150 (0.44 percent to 1.50 percent) for the experimental sets in Tables D-1, D-2, D-3, D-4, and D-5, respectively. No particular correlation seems to exist between WCD and the lattice configuration or pitch. Neither does there seem to be a correlation with plutonium composition type. The plutonium composition types are given in Table B-1 of Appendix B to this SRP and are designated as weapons grade (WG), fuel grade (FG), and power grade (PG).
The maximum value for WCD found in the five tables is 0.0328, or 3.28 percent in keff. How accurately a criticality computer code can predict the critical value for a criticality experiment depends on the methodology employed by the code and the cross-section set used, together with the detail to which the experimental system is modeled in the input to the computer code.
In addition, the basic experimental uncertainty limits the ultimate prediction accuracy possible.
Of particular importance is the cross-section set. Values for WCD in the five tables that are significantly less than 0.0100 are due to the fact that kcalc is greater than kexp. Therefore, the value for Eff-Bias, in that case, is just the negative of twice the calculated standard deviation, which is approximately 0.0020. The cross-section sets used in the analyses represented in the tables over-predict plutonium reactivity, and this represents some of the reason for the over-prediction for kcalc for these experiments. Values for ksafe are not expected to be much above 0.93, except when it can be demonstrated that the criticality code and cross section set overestimate the reactivity of the MOX contents.
Analyzing an acceptable number of MOX benchmarks is the preferred way to obtain a bias value for the MOX contents of a package. With the relatively limited number of MOX critical experiments available for use in validation exercises, it is important to determine that the application of interest to the reviewer fits within the area of applicability for the set of critical benchmark experiments selected for validation. Guidance on how to select an appropriate set of benchmark experiments for a fissile system is given in NUREG/CR-5661 and in NUREG/CR-6361. A computational methodology to select an appropriate set of benchmark experiments for a fissile package application has also been developed for SCALE (Broadhead et al. 1999; Broadhead et al. 2004; Rearden and Childs 2000; Rearden and Mueller 2008; Dunn and Rearden 2001).
Beginning with version 5 of SCALE, a set of sensitivity and uncertainty analysis tools have been developed and are included with the code that gives a measure of the similarity of the reactivity of a package application to that of an experimental benchmark. Successive versions of SCALE include an improved and expanded set of tools (Perfetti and Rearden 2016; Rearden et al. 2011; Perfetti et al. 2016; Williams et al. 2013, ORNL 2011). Sensitivity coefficients for both systems are computed and give the sensitivity of each systems keff to the cross-section 2
If the benchmarks are applied to a package application where there is a lack of experimental data, the 0.05 administrative margin may not be sufficient, and the reviewer needs to be aware of this issue. In reality, the 0.05 margin should be sufficient, but there needs to be an assessment of the adequacy of the 0.05 to establish the basis. Guidance for deciding on an acceptable choice for the administrative margin is given in NUREG/CR-5661. See also NUREG/CR-6361.
D-3
data. These sensitivity coefficients are determined for each energy group in the cross-section library chosen in the analysis, as well as the sum over all energy groups. Two integral parameters for the combined systems are produced from the sensitivity data to determine system-to-system similarities. The first parameter can be used as a gauge of system similarity to sensitivity only. The second parameter can be used as a measure of the similarity of the systems in terms of uncertainty, not just sensitivity. The pair of integral parameter values is determined for every potential benchmark experiment with the package application of interest.
When two systems produce an appropriately high value (i.e., a value sufficiently close to 1) for either integral parameter, or both, this indicates the keff response is similar enough that one system serves well to validate the criticality safety parameters for the other system. Previous analyses using these tools have used the value of 0.8 as a threshold for determining that systems under consideration are similar enough; this is consistent with recommendations the SCALE developer, Oak Ridge National Laboratory, has made. The benchmark experiments chosen for complete validation are those with high integral parameter values (Broadhead et al.
1999; Broadhead et al. 2004; Rearden and Childs 2000; Rearden and Mueller 2008; Dunn and Rearden 2001).
New parameters can also be constructed from the components of the integral parameters and can be used to explore the sensitivity of specific nuclide reactions of benchmark experiments with the package application of interest. For example, if low integral parameter values are found for an application with all benchmark experiments chosen for validation, the new parameters could serve to identify which nuclides would require additional experimental benchmark data for complete validation. Also, in the validation of transportation packages for commercial fuel, numerous benchmark experiments might serve to validate the fission reactions, and thus high integral parameter values would be found. However, the new parameters could be used to find benchmarks to ensure that any poison materials in the package are also well validated by the benchmarks. With the inclusion of these sensitivity and uncertainty analysis tools in the SCALE code, beginning with version 5, the criticality safety analyst now has a powerful set of tools available to perform detailed quantitative analyses to determine the applicability of benchmark experiments to help design package applications under consideration (Broadhead et al. 1999; Broadhead et al. 2004; Rearden and Childs 2000; Rearden and Mueller 2008; Dunn and Rearden 2001).
D-4
Table D-1 Important characteristics of lattice experiments with weight percent of Pu/(Pu+U) to 5 percent (from IHECSBE)
Designation for experiments a MCT-009 MCT-002 MCT-002 MCT-006 MCT-007 MCT-008 MCT-004 MCT-005 Facility where experiments conducted Hanford Hanford Hanford Hanford Hanford Hanford Tokai Hanford Computer codes used in evaluations b MCNP/KENO MCNP MCNP MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO Cross section sets used in evaluations c ENDF/B-V/IV ENDF/B-V ENDF/B-V ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV JENDL-3.2 ENDF/B-V/IV Cross section type d cont/27grp cont cont cont/27grp cont/27grp cont/27grp cont/137grp cont/27grp Fuel compound e oxide oxide oxide oxide oxide oxide oxide oxide Fuel compound form solid solid solid solid solid solid solid solid Density of fuel f 86.7% 86.7% 86.7% 86.7% 86.7% 86.7% 55% 86%
Organization of fuel g pins pins pins pins pins pins pins pins Cladding used for fuel h Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 Zirc-2 Pu/(Pu+U) atom percent 1.51% 1.80% 1.80% 1.80% 2.01% 2.01% 3.03% 3.52%
235 U atom percent 0.16% 0.71% 0.71% 0.71% 0.72% 0.72% 0.71% 0.71%
238 U atom percent 99.84% 99.29% 99.29% 99.29% 99.28% 99.28% 99.29% 99.29%
238 Pu atom percent - 0.01% 0.01% 0.01% - - 0.50% 0.28%
239 Pu atom percent 91.41% 91.84% 91.84% 91.84% 81.11% 71.76% 68.18% 75.39%
240 Pu atom percent 7.83% 7.76% 7.76% 7.76% 16.54% 23.50% 22.02% 18.10%
241 Pu atom percent 0.73% 0.37% 0.37% 0.37% 2.15% 4.08% 7.26% 5.08%
242 Pu atom percent 0.03% 0.03% 0.03% 0.03% 0.20% 0.66% 2.04% 1.15%
Plutonium type as given in Table B-1 WG WG WG WG FG PG PG FG-PG D-5 Shape of lattice i cylinder rectangle rectangle cylinder cylinder cylinder rectangle cylinder Pitch of lattice triangle square square triangle triangle triangle square triangle Number of experiments in each set 6 3 3 6 5 6 4 7 Fissile moderator used j H2O H2O B-H2O H2O H2O H2O H2O H2O Reflector used H2O H2O B-H2O H2O H2O H2O H2O H2O Maximum effective bias of experiments in set -0.0112 -0.0052 -0.0026 -0.0089 -0.0040 -0.0068 -0.0097 -0.0037 (Eff-Bias)
Maximum uncertainty of experiments in set 0.0080 0.0059 0.0045 0.0054 0.0061 0.0065 0.0051 0.0042 (Exp-Uncer)
Exp-Uncer minus Eff-Bias (WCD) 0.0192 0.0111 0.0071 0.0143 0.0101 0.0133 0.0148 0.0079 a
MCT = MIX-COMP-THERM.
b Codes MCNP (LANL, 1997) and KENO (ORNL, 1995).
c ENDF/B-V/IV means cross section set ENDF/B-V for MCNP and cross section set ENDF/B-IV for KENO. JENDL-3.2 is the cross section set for both MCNP and KENO.
d Cross section type is either continuous cross sections (cont.) or group cross sections (27grp, 137grp).
e Heavy metal is as an oxide.
MOX density given as percent of theoretical density taken as 11.00 g/cm3.
g Pins means organization of MOX is as pellets in fuel pins.
h Zirc-2 means Zircaloy-2 cladding.
I Cylinder means shape of lattice is a cylinder. Rectangle means shape of lattice is a rectangle.
j B-H2O means borated water as moderator or reflector.
f
Table D-2 Important characteristics of lattice experiments with weight percent of Pu/(Pu+U) from 5 percent to 15 percent (from IHECSBE)
Designation for experiments a MCT-003 MCT-003 MCT-012 MCT-012 MCT-012 MCT-012 Facility where experiments conducted WREC WREC Hanford Hanford Hanford Hanford Computer codes used in evaluations b MCNP MCNP MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO Cross section sets used in evaluations c ENDF/B-V ENDF/B-V ENDF/B-V ENDF/B-V ENDF/B-V ENDF/B-V Cross section type d cont cont cont/238grp cont/238grp cont/238grp cont/238grp Fuel compound e oxide oxide oxide-poly oxide-poly oxide-poly oxide-poly Fuel compound form solid solid solid solid solid solid Density of fuel f 94% 94% N/A N/A N/A N/A Organization of fuel g pins pins cubes, slabs cubes, slabs cubes, slabs cubes, slabs Cladding used for fuel h Zirc-4 Zirc-4 plastic 471 plastic 471 plastic 471 plastic 471 Pu/(Pu+U) atom percent 6.63% 6.63% 7.60% 7.89% 14.62% 14.62%
235 U atom percent 0.71% 0.71% 0.15% 0.15% 0.15% 0.15%
238 U atom percent 99.29% 99.29% 99.85% 99.85% 99.85% 99.85%
238 Pu atom percent - - 0.59% - - -
239 Pu atom percent 90.65% 90.65% 67.97% 91.25% 91.42% 91.42%
240 Pu atom percent 8.55% 8.55% 22.95% 8.12% 7.97% 7.97%
241 Pu atom percent 0.76% 0.76% 5.57% 0.58% 0.57% 0.57%
242 Pu atom percent 0.04% 0.04% 2.92% 0.05% 0.04% 0.04%
Plutonium type as given in Table B-1 WG-FG WG-FG PG WG WG WG Shape of lattice i rectangle rectangle 3D cube 3D cube 3D cube 3D cube Pitch of lattice square square square square square square D-6 Number of experiments in each set 5 1 6 7 6 3 Fissile moderator used j H2O B-H2O polystyrene polystyrene polystyrene polystyrene Reflector used H2O B-H2O Plexiglas Plexiglas Plexiglas none Maximum effective bias of experiments in set (Eff-
-0.0063 -0.0030 -0.0270 -0.0016 -0.0016 -0.0020 Bias)
Maximum uncertainty of experiments in set (Exp-0.0071 0.0052 0.0058 0.0036 0.0027 0.0037 Uncer)
Exp-Uncer minus Eff-Bias (WCD) 0.0134 0.0082 0.0328 0.0052 0.0043 0.0057 a
MCT = MIX-COMP-THERM.
b Codes MCNP (LANL, 1997) and KENO (ORNL, 1995).
c ENDF/B-V is the cross section set for MCNP and KENO.
d Cross section type is either continuous cross sections (cont) or group cross sections (238grp).
e Heavy metal is as an oxide. Oxide-poly means mixture of MOX particles and polystyrene pressed into cubes and slabs.
f MOX density given as percent of theoretical density taken as 11.00 g/cm3.
g Pins means organization of MOX is as pellets in fuel pins. Cubes, slabs means organization of MOX-polystyrene is as cubes and slabs.
h Zirc-4 means Zircaloy-4 cladding. Plastic 471 means cladding is six mil plastic tape MM&M (3M) #471.
I Rectangle means shape of lattice is a rectangle. 3D cube means cubes and slabs stacked into the shape of a 3D rectangular cube.
j B-H2O means borated water as moderator or reflector.
Table D-3 Important characteristics of lattice experiments with weight percent of Pu/(Pu+U) greater than 15 percent (from IHECSBE)
Designation for experiments a MCT-001 MCT-011 MCT-012 MCT-012 Facility where experiments conducted Hanford Valduc Hanford Hanford Computer codes used in evaluations b MONK MORET MCNP/KENO MCNP/KENO Cross section sets used in evaluations c UKNDL JEF2.2 ENDF/B-V ENDF/B-V Cross section type d cont 172gp cont/238grp cont/238grp Fuel compound e oxide oxide oxide-poly oxide-poly Fuel compound form solid solid solid solid Density of fuel f 89.4% 94.2% N/A N/A Organization of fuel g pins pins cubes, slabs cubes, slabs Cladding used for fuel h 316SS Z3CND18.12 SS plastic 471 plastic 471 Pu/(Pu+U) atom percent 19.70% 25.80% 30.00% 30.00%
235 U atom percent 0.71% 60.15% 0.15% 0.15%
238 U atom percent 99.29% 39.85% 99.85% 99.85%
238 Pu atom percent 0.15% - - -
239 Pu atom percent 85.54% 89.00% 91.22% 91.22%
240 Pu atom percent 11.46% 9.72% 8.13% 8.13%
241 Pu atom percent 2.50% 1.21% 0.61% 0.61%
242 Pu atom percent 0.35% 0.07% 0.04% 0.04%
Plutonium type as given in Table B-1 FG WG-FG WG WG Shape of lattice i rectangle cylinder 3D cube 3D cube D-7 Pitch of lattice square triangle square square Number of experiments in each set 4 6 8 3 Fissile moderator used H2O H2O polystyrene polystyrene Reflector used H2O H2O Plexiglas none Maximum effective bias of experiments in set (Eff-Bias) -0.0103 -0.0006 -0.0018 -0.0086 Maximum uncertainty of experiments in set (Exp-Uncer) 0.0025 0.0017 0.0049 0.0052 Exp-Uncer minus Eff-Bias (WCD) 0.0128 0.0023 0.0067 0.0138 a
MCT = MIX-COMP-THERM.
b Codes MCNP (LANL, 1997), KENO (ORNL, 1995), MONK, and MORET. MONK is a three-dimensional Monte Carlo radiation transport code that uses point-wise cross sections, developed by A.E.A. Technology of the United Kingdom. MORET is a three-dimensional Monte Carlo criticality code that uses multigroup cross sections, developed by C.E.A. of France.
c ENDF/B-V is the cross section set for MCNP and KENO. UKNDL is the cross section set for MONK. JEF2.2 is the cross section set for MORET.
d Cross section type is either continuous cross sections (cont) or group cross sections (172grp, 238grp).
e Heavy metal is as an oxide. Oxide-poly means mixture of MOX particles and polystyrene pressed into cubes and slabs.
f MOX density given as percent of theoretical density taken as 11.00 g/cm3.
g Pins means organization of MOX is as pellets in fuel pins. Cubes, slabs means organization of MOX-polystyrene is as cubes and slabs.
h SS means stainless steel cladding. Plastic 471 means cladding is six mil plastic tape MM&M (3M) #471.
I Cylinder means shape of lattice is a cylinder. Rectangle means shape of lattice is a rectangle. 3D cube means cubes and slabs stacked into the shape of a 3D rectangular cube.
Table D-4 Important characteristics of tank experiments with weight percent of Pu/(Pu+U) to 31 percent (from IHECSBE)
Designation for experiments a MST-001 MST-001 MST-001 MST-001 MST-002 MST-003 Facility where experiments conducted Hanford Hanford Hanford Hanford Hanford AWRE Computer codes used in evaluations b MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MONK Cross section sets used in evaluations c ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV UKNDL Cross section type d cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp cont Fuel compound e nitrate nitrate nitrate nitrate nitrate nitrate Fuel compound form liquid liquid liquid liquid liquid liquid Density of fuel f 1.31-1.68 1.31-1.68 1.31-1.48 1.70 1.09 1.11-1.52 Pu/(Pu+U) atom percent 22% 22% 22% 22% 23% 30.7%
235 U atom percent 0.70% 0.70% 0.70% 0.70% 0.70% 0.72%
238 U atom percent 99.30% 99.30% 99.30% 99.30% 99.30% 99.28%
238 Pu atom percent 0.03% 0.03% 0.03% 0.03% 0.03% -
239 Pu atom percent 91.12% 91.12% 91.12% 91.12% 91.12% 93.95%
240 Pu atom percent 8.34% 8.34% 8.34% 8.34% 8.31% 5.63%
241 Pu atom percent 0.42% 0.42% 0.42% 0.42% 0.45% 0.42%
242 Pu atom percent 0.09% 0.09% 0.09% 0.09% 0.09% -
Plutonium type as given in Table B-1 WG WG WG WG WG WG Tank fissile liquid is in g N/A cylinder cylinder cylinder cylinder slab Auxiliary tank additional fissile liquid is in h annular annular annular N/A N/A N/A Number of experiments in each set 2 5 2 1 1 10 Fissile moderator used i soln H2O soln H2O soln H2O soln H2O soln H2O soln H2O D-8 Reflector used j B4C-concrete B4C-concrete poly-Cd cover none H2O H2O & poly Maximum effective bias of experiments in set (Eff-
-0.0101 -0.0164 -0.0028 -0.0068 -0.0020 -0.0038 Bias)
Maximum uncertainty of experiments in set (Exp-0.0016 0.0016 0.0016 0.0016 0.0024 0.0025 Uncer)
Exp-Uncer minus Eff-Bias (WCD) 0.0117 0.0180 0.0044 0.0084 0.0044 0.0063 a
MST = MIX-SOL-THERM.
b Codes MCNP (LANL, 1997), KENO (ORNL, 1995), and MONK (A.E.A. Technology). MONK is a three-dimensional Monte Carlo radiation transport code that uses point-wise cross sections, developed by A.E.A. Technology of the United Kingdom.
c ENDF/B-V/IV means cross section set ENDF/B-V for MCNP and ENDF/B-IV for KENO. UKNDL is cross section set for MONK.
d Cross section type is either continuous cross sections (cont) or group cross sections (27grp).
e Heavy metal is as a nitrate dissolved in dilute nitric acid solution.
f Solution density is in g/ml.
g Containers for fissile solution are cylinders or slabs.
h Annular tank surrounding central cylindrical tank or just an annular tank.
I Soln H2O means the moderator is the fissile nitrate solution.
j B4C-concrete means borated concrete. Poly-Cd cover means polyethylene reflector coated with Cd.
Table D-5 Important characteristics of tank experiments with weight percent of Pu/(Pu+U) greater than 31 percent (from IHECSBE)
Designation for experiments a MST-004 MST-004 MST-004 MST-005 MST-005 MST-002 MST-001 Facility where experiments conducted Hanford Hanford Hanford Hanford Hanford Hanford Hanford Computer codes used in evaluations b MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO MCNP/KENO Cross section sets used in evaluations c ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV ENDF/B-V/IV Cross section type d cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp cont/27grp Fuel compound e nitrate nitrate nitrate nitrate nitrate nitrate nitrate Fuel compound form liquid liquid liquid liquid liquid liquid liquid Density of fuel f 1.17-1.67 1.17-1.67 1.17-1.67 1.17-1.67 1.17-1.67 1.05 1.15-1.44 Pu/(Pu+U) atom percent 40% 40% 40% 40% 40% 52% 97%
235 U atom percent 0.56% 0.56% 0.56% 0.56% 0.56% 0.70% 2.29%
238 U atom percent 99.44% 99.44% 99.44% 99.44% 99.44% 99.30% 97.71%
238 Pu atom percent 0.03% 0.03% 0.03% 0.03% 0.03% 0.03% 0.03%
239 Pu atom percent 91.12% 91.12% 91.12% 91.12% 91.12% 91.12% 91.57%
240 Pu atom percent 8.34% 8.34% 8.34% 8.34% 8.34% 8.34% 7.94%
241 Pu atom percent 0.42% 0.42% 0.42% 0.42% 0.42% 0.42% 0.39%
242 Pu atom percent 0.09% 0.09% 0.09% 0.09% 0.09% 0.09% 0.07%
Plutonium type as given in Table B-1 WG WG WG WG WG WG WG Tank fissile liquid is in g cylinder cylinder cylinder slab slab cylinder cylinder D-9 Auxiliary tank additional fissile liquid is in h N/A N/A N/A N/A N/A N/A annular Number of experiments in each set 3 3 3 3 4 2 3 Fissile moderator used i soln H2O soln H2O soln H2O soln H2O soln H2O soln H2O soln H2O Reflector used j none H2O concrete none H2O H2O B4C-concrete Maximum effective bias of experiments in set (Eff-Bias) -0.0060 -0.0048 -0.0024 -0.0114 -0.0026 -0.0020 -0.0032 Maximum uncertainty of experiments in set (Exp-Uncer) 0.0033 0.0033 0.0078 0.0036 0.0037 0.0024 0.0016 Exp-Uncer minus Eff-Bias (WCD) 0.0093 0.0081 0.0102 0.0150 0.0063 0.0044 0.0048 a
MST = MIX-SOL-THERM.
b Codes MCNP (LANL, 1997) and KENO (ORNL, 1995).
c ENDF/B-V/IV means ENDF/B-V for MCNP and ENDF/B-IV for KENO.
d Cross section type is either continuous cross sections (cont) or group cross sections (27grp).
e Heavy metal is as a nitrate dissolved in dilute nitric acid solution.
f Solution density is in g/ml.
g Containers for fissile solution are cylinders or slabs.
h Annular tank surrounding central cylindrical tank.
i Soln H2O means the moderator is the fissile nitrate solution.
j B4C-concrete means borated concrete.
D.3 References Broadhead, B.L., et al., Sensitivity and Uncertainty Analyses Applied to Criticality Safety Validation, NUREG/CR-6655, Vols. 1 and 2 (ORNL/TM-13692/V1 and V2), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, November 1999.
Broadhead, B.L., et al., Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques, Nuclear Science and Engineering, 146:3, 340-366 (2004).
Dunn, M.E., and B.T. Rearden, Application of Sensitivity and Uncertainty Analysis Methods to a Validation Study for Weapons-Grade Mixed-Oxide Fuel, 2001 ANS Embedded Topical Meeting on Practical Implementation of Nuclear Criticality Safety, Reno, NV, November 11-15, 2001.
Dyer, H.R., and C.V. Parks, Recommendations for Preparing the Criticality Safety Evaluation of Transportation Packages, NUREG/CR-5661 (ORNL/TM-11936), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, April 1997.
Fernex, F., Programme HTC-Phase 1: Réseaux de Crayons dans lEau Pure (Water-Moderated and Reflected Simple Arrays) Reevaluation des Expériences, DSU/SEC/T/2005-33/D.R.,
Valduc, France, IRSN (2006). PROPRIETARY document.
Fernex, F., Programme HTC-Phase 2: Réseaux Simples en Eau Empoisonnée (Bore et Gadolinium) (Reflected Simple Arrays Moderated by Poisoned Water with Gadolinium or Boron)
Réévaluation des Experiences, DSU/SEC/T/2005-38/D.R., Valduc, France, IRSN (2006).
PROPRIETARY document.
Fernex, F., Programme HTC-Phase 3: Configurations Stockage en Piscine (Pool Storage)
Réévaluation des Experiences, DSU/SEC/T/2005-37/D.R., Valduc, France, IRSN (2006).
PROPRIETARY document.
Fernex, F., Programme HTC-Phase 4: Configurations Châteaux de Transport (Shipping Cask)
Réévaluation des Experiences, DSU/SEC/T/2005-36/D.R., Valduc, France, IRSN (2006).
PROPRIETARY document.
Lichtenwalter, J.J., et al., Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, NUREG/CR-6361 (ORNL/TM-13211), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, March 1997.
Los Alamos National Laboratory (LANL), MCNP - A General Monte Carlo N-Particle Transport Code, Version 4B, Judith F. Briesmeister, Editor, Los Alamos National Laboratory, LA-12625-M, March 1997. Various versions of MCNP are available; this reference is for version 4B.
Mueller, D.E., K.R. Elam, and P.B. Fox, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, NUREG/CR-6979 (ORNL/TM-2007/083), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, TN, September 2008.
Nuclear Energy Agency (NEA), International Handbook of Evaluated Criticality Safety Benchmark Experiments, Organization for Economic Co-operation and Development, NEA/NSC/DOC(95)03, September 2001 Edition.
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Nuclear Energy Agency (NEA), International Handbook of Evaluated Criticality Safety Benchmark Experiments, Organization for Economic Co-operation and Development, NEA/NSC/DOC(95)03, September 2014 Edition.
Oak Ridge National Laboratory (ORNL), SCALE: A Modular Code System for Performing Standardized Analyses for Licensing Evaluations, NUREG/CR-0200, Rev. 4 (ORNL/NUREG/CSD-2/R4), Vols. I, II, III, October 1995. Various versions of SCALE are available; this reference is for version 4.3.
Oak Ridge National Laboratory (ORNL), Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, TN, June 2011. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-785.
Perfetti, C.M., & B.T. Rearden, Development of a Generalized Perturbation Theory Method for Sensitivity Analysis Using Continuous-Energy Monte Carlo Methods, Nuclear Science and Engineering, 182:3, 354-368 (2016).
Perfetti, C.M., B.T. Rearden, and W.R. Martin, SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations, Nuclear Science and Engineering, 182:3, 332-353 (2016).
Rearden, B.T., and R.L. Childs, Prototypical Sensitivity and Uncertainty Analysis Codes for Criticality Safety with the SCALE Code System, Trans. Am. Nucl. Soc., Washington, DC, November 2000.
Rearden, B.T., and D.E. Mueller, Recent Use of Covariance Data for Criticality Safety Assessment, Nuclear Data Sheets, 109, 2739-2744 (2008).
Rearden, B.T., et al., Sensitivity and Uncertainty Analysis Capabilities and Data in SCALE, Nuclear Technology, 174:2, 236-288 (2011).
Williams, M.L., et al., A Statistical Sampling Method for Uncertainty Analysis with SCALE and XSUSA, Nuclear Technology, 183:3, 515-526 (2013).
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