Technical support for a proposed decay heat guide using SAS2H/ORIGEN-S data

Hermann, O.W.; Parks, C.V.; Renier, J.P.

doi:10.2172/43752

Cited by 11 publications

(5 citation statements)

References 1 publication

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“…The structure material compositions include all materials consisting fuel assemblies except heavy nuclides in the fuel pellets. Those data were taken from References [13,[16][17][18]. For the PWR fuel assemblies, the library PWR34J40 (PWR, 17×17 fuel assembly, 235 U enrichment 3.4 wt%) or PWR41J40 (PWR, 17×17 fuel assembly, 235 U enrichment 4.1 wt%) [3] was used depending on the assembly average enrichments of the fuel assemblies.…”

Section: Origen22 Calculation Conditions and Resultsmentioning

confidence: 99%

“…In the calculations with CASMO5, the threedimensional fuel assemblies were represented by two-dimensional assembly models. The detailed design specifications of the two-dimensional fuel assembly models and the operating histories of thermal hydraulic parameters such as fuel temperatures of fuel pellets, fuel claddings and moderator, and moderator densities were taken from References [13,[16][17][18]. The inputs of structure materials were set in the calculations; however, nuclear transmutation of the nuclides in the structure material is not fully treated for decay heat calculations of CASMO5.…”

Section: Casmo5 Calculation Conditions and Resultsmentioning

confidence: 99%

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Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Yamamoto¹,

Iwahashi²

2016

Journal of Nuclear Science and Technology

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Comparing with the fission product nuclide (FP) decay heat summation calculation result in MeV/sec/fission based on the JENDL FP decay and yield data files 2011 for the burst fission, FP decay heat calculated by ORIGEN2.2 coupled with JENDL-4.0 base library ORLIBJ40 was verified at the cooling time from 1 sec to 10 8 sec for 235 U (thermal), 238 U (fast), 239 Pu (thermal) and 241 Pu (thermal). For these fission nuclides, FP decay heat calculated by CASMO5 at the same cooling time after a short irradiation (10 4 sec) was also compared with that of ORIGEN2.2. In the analysis of decay heat measurements at the cooling time from 2.3 years to 27 years consisting of four data sets on the fuel assemblies discharged from the US PWRs and BWRs, and the Swedish PWRs and BWRs, the average values of the ratios of the calculated to measured results (C/E's) were from 0.972 to 1.031 for ORIGEN2.2, and from 0.977 to 1.016 for CASMO5. The standard deviations of C/E's for the four data sets were from 0.02 to 0.03 for the both codes except for those of the US BWR fuel assemblies which were from 0.11 to 0.12. The obtained C/E's were similar to those in the precedent study.

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Section: Origen22 Calculation Conditions and Resultsmentioning

confidence: 99%

Section: Casmo5 Calculation Conditions and Resultsmentioning

confidence: 99%

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Yamamoto¹,

Iwahashi²

2016

Journal of Nuclear Science and Technology

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“…The uniform distribution of decay heat is assumed within the range of 98% and 102% of the base value. The reasons for the power variations are the uncertainties related to the calorimetric error when measuring thermal power, and the standard deviations in fission product yields, half-lives and the amount of energy released during the radioactive decay [37]. The generation of heat has a key impact on the thermal response of the core, its damage and fuel rod cladding oxidation, which exponentially depends on temperature.…”

Section: Selection Of Uncertain Parametersmentioning

confidence: 99%

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

Šadek

Grgić

Allison³

et al. 2022

Energies

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A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.

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“…The initial uranium content of each group was determined from the initial 235 U enrichment of the associated fuel batch. The following empirical relationship was used to determine relative isotopic content: 18 w 234 = 0.007731(w 235 ) 1.0837 , w 236 = 0.0046w 235 ,…”

Section: Sas2h Fuel Groupsmentioning

confidence: 99%

Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

Bowman¹

1995

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The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k eff) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4 SAS2H sequence. These reactor critical benchmarks have been reanalyzed using the methodology described above. The two benchmark critical calculations were the beginning-of-cycle (BOC) startup at hot, zero-power (HZP) and an end-of-cycle (EOC) critical at hot, full-power (HFP) critical conditions. These calculations were used to check for consistency in the calculated results for different burnup, downtime, temperature, xenon, and boron conditions. The k eff results were 1.0014 and 1.0113, respectively, with a standard deviation of 0.0005. viii

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Technical support for a proposed decay heat guide using SAS2H/ORIGEN-S data

Cited by 11 publications

References 1 publication

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

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