• Energy Research

Abstract A photon transport capability has been implemented and verified in the Monte Carlo code MCS of Ulsan National Institute of Science and Technology for the purpose of radiation shielding studies. The MCS photon fixed-source mode simulates the transport of photons between 1 keV and 20 MeV for all elements from hydrogen to fermium. The specific physics for the main four photo-atomic reactions (Rayleigh scattering, Compton scattering, photoelectric effect and pair production) and three secondary processes of photon production (positron–electron annihilation, atomic relaxation and electron/positron bremsstrahlung) are reviewed. Verification results against Monte Carlo codes MCNP6.1 and SERPENT2.1.29 are presented. The verification cases include the comparison of energy distributions of photon flux in an infinite medium, of spatial distributions of photon flux in a cylinder, of the spatial distribution of photon body-equivalent dose in a spent nuclear fuel transport cask, and of photon KERMA (Kinetic Energy Released per MAss) in photon detector calibration geometries. Good calculation/calculation agreement is observed overall, with some marked differences in the detailed photon flux comparison at given energy ranges traced back to differences in photon physics implementation. MCS can from now on be applied for the purpose of advanced photon studies and corresponding validation against experimental shielding benchmarks will follow in the future.

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as 241Am, 137Ba, 244Cm, 238Pu, and 90Y.

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol’ indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

Abstract This paper presents a study of uncertainty quantification (UQ) of spent nuclear fuel isotope inventory. Takahama-3 realistic assays were used for calculation to quantify the impact of three different parameter groups, namely, design parameters (i.e., fuel pellet radius, clad outer radius, UO2 density, and UO2 enrichment), operation conditions (i.e., power density, fuel temperature, moderator temperature, and boron concentration), and covariance of nuclear data. The UQ study is focused on analysis of fission products and actinides used in criticality analysis during the back-end cycle. Stochastic sampling method was used for the UQ and Lagrange interpolation is used for the prediction of isotope inventory. In-house two step code system, STREAM/RAST-K was used for calculation with ENDF/B-VII.1 cross section library and ENDF/B-VII.0 decay library. Most of the actinides and fission products are calculated within ±5% of experiment. In addition, 238U and 239Pu have the largest contribution on keff uncertainty. Moreover, 245Cm and 246Cm have the largest isotope inventory uncertainty compared with other isotopes and the covariance of nuclear data has the most dominant effect on the uncertainty of isotope composition.

Abstract This paper presents the verification and validation of the radiation source term calculation capability implemented in the reactor analysis code STREAM for pressurized water reactor (PWR) spent nuclear fuel (SNF) analysis. Activity, decay heat, neutron and gamma source spectra of irradiated PWR fuel assemblies are calculated with STREAM and ORIGEN and the results obtained are compared. The verification is performed with Westinghouse 15 × 15, 17 × 17 and Combustion Engineering 16 × 16 designs with enrichment of 3–5 wt.% 235U, burnup of 30–60 GWd/t and cooling times of 0.3–1000 years. The radiation source term comparison between STREAM and ORIGEN indicates that the implementation of the source term equations is correct. The validation is done by comparing STREAM calculation results against 91 decay heat calorimetric measurements for 52 PWR fuel assemblies performed at the Swedish central storage facility for spent nuclear fuel, CLAB, and United States Hanford Engineering Development Laboratory (HEDL) and General Electric Morris Operations. The measured fuel assemblies have enrichment of 2–4 wt.% 235U, burnup of 28–51 GWd/t and cooling times of 2–27 years. The comparison between calculated and measured decay heat shows good agreement. For most of the measurements analyzed, the calculated decay heat is within the range of the measurement uncertainty. The average, overall C/E over 71 CLAB decay heat measurements is 1.000 ± 0.017, and 0.997 ± 0.013 for 20 US measurements. Overall, this work demonstrates that STREAM can be reliably used to predict PWR SNF radiation source terms.

Abstract This work investigates the depletion capability implemented in lattice physics code STREAM for the prediction of pressurized water reactor (PWR) uranium dioxide (UO2) spent nuclear fuel (SNF) isotopic inventory. The validation of this capability is performed by comparison of STREAM calculation results to measured SNF assay data obtained from PWRs Takahama-3, Calvert Cliffs and GKN II. The depletion analysis is conducted with the ENDF/B-VII.0 library and uses a pin cell model of the fuel rods from which the fuel samples were taken. The Chebyshev Rational Approximation Method (CRAM) is used to solve the depletion equation with about 1300–1600 isotopes in the depletion chain. 16 actinides and 23 fission products are analyzed in 14 spent UO2 fuel samples. The actinides are isotopes of uranium, neptunium, plutonium, americium and curium. The fission products nuclides include isotopes of cesium, neodymium, europium, samarium as well as 106Ru, 144Ce, 155Gd, 99Tc, 90Sr, 109Ag, and 103Rh. The sensitivity of some of the nuclides to the details of the power history and the adjustment of the fuel sample burnup is discussed. The impact of using ENDF/B-VII.0 library instead of ENDF/B-VI.8 is also discussed. Most of the nuclides analyzed are well predicted within ±7% of the experiment for actinides and fission products. STREAM depletion results are also compared to the codes SWAT, HELIOS and SCALE results based on publicly available information in literature, to check the performance of STREAM relative to other codes for the prediction of SNF isotopic inventory. The comparison to other code systems shows that the implementation in STREAM is of comparable accuracy. Overall, this paper demonstrates that the depletion capability in STREAM can be reliably applied to predict the isotopic inventory of PWR UO2 SNF for burnup ranging from 14 to 54 GWd/t and initial enrichment ranging from 3.0 to 4.1 wt% 235U.