Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat

The decay heat rate of five spent nuclear fuel assemblies of the pressurized water reactor type were measured by calorimetry at the interim storage for spent nuclear fuel in Sweden. Calculations of the decay heat rate of the five assemblies were performed by 20 organizations using different codes and nuclear data libraries resulting in 31 results for each assembly, spanning most of the current state-of-the-art practice. The calculations were based on a selected subset of information, such as reactor operating history and fuel assembly properties. The relative difference between the measured and average calculated decay heat rate ranged from 0.6% to 3.3% for the five assemblies. The standard deviation of these relative differences ranged from 1.9% to 2.4%

ADVANCES IN STUDSVIK’S SYSTEM FOR SPENT FUEL ANALYSES

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations, fluxes, and cross-sections, calculated by the CASMO5 neutron transport and depletion code, with irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the SNF code to compute the spent nuclear fuel characteristics. Recent advances in the system, including cross-sections and decay data from ENDF/B-VIII.R0, are presented in this paper, together with validation results against decay heat power and isotopic compositions measurements. Measurements conducted at the Swedish interim storage facility, CLAB, are used for validation of the decay heat power, while comparisons to the results of the international program ARIANE are used to demonstrate the capability of CMS5/SNF to accurately predict isotopic compositions. The paper shows the results calculated with ENDF/B-VIII.R0, and the effect on the spent fuel characteristics is evaluated by comparisons to the earlier ENDF/B-VII.R1 results

Update and evaluation of decay data for spent nuclear fuel analyses

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data

ADVANCES IN STUDSVIK’S SYSTEM FOR SPENT FUEL ANALYSES

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations, fluxes, and cross-sections, calculated by the CASMO5 neutron transport and depletion code, with irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the SNF code to compute the spent nuclear fuel characteristics. Recent advances in the system, including cross-sections and decay data from ENDF/B-VIII.R0, are presented in this paper, together with validation results against decay heat power and isotopic compositions measurements. Measurements conducted at the Swedish interim storage facility, CLAB, are used for validation of the decay heat power, while comparisons to the results of the international program ARIANE are used to demonstrate the capability of CMS5/SNF to accurately predict isotopic compositions. The paper shows the results calculated with ENDF/B-VIII.R0, and the effect on the spent fuel characteristics is evaluated by comparisons to the earlier ENDF/B-VII.R1 results

Update and evaluation of decay data for spent nuclear fuel analyses

Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data

Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat

This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization

Analysis of ENRESA BWR samples: nuclide inventory and decay heat

In this paper the isotopic compositions from 8 Boiling Water Reactor samples are analyzed following different irradiation assumptions as well as different simulation tools. These samples are part of a proprietary experimental program by a Spanish consortium, and they were obtained from a GE14 assembly irradiated in Sweden. Calculated nuclide concentrations are compared with measured ones providing biases for a selection of isotopes and samples; calculated uncertainties are also provided. Finally, the decay heat from one the sample segment is calculated and compared among the different simulation assumptions. It is shown that depending on the considered nuclear data library and modeling, different contributors affect the calculated quantities, indicating a certain level of prediction power

Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat

International audienceThe decay heat rate of five spent nuclear fuel assemblies of the pressurized water reactor type were measured by calorimetry at the interim storage for spent nuclear fuel in Sweden. Calculations of the decay heat rate of the five assemblies were performed by 20 organizations using different codes and nuclear data libraries resulting in 31 results for each assembly, spanning most of the current state-of-the-art practice. The calculations were based on a selected subset of information, such as reactor operating history and fuel assembly properties. The relative difference between the measured and average calculated decay heat rate ranged from 0.6% to 3.3% for the five assemblies. The standard deviation of these relative differences ranged from 1.9% to 2.4%

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors : a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier