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The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous ap...

The ITER fusion reactor is being built to demonstrate the feasibility of fusion power and will be the largest tokamak in the world. The tokamak cooling water system (TCWS) will extract the ...

Heat generated in the ITER fusion reactor is deposited in the tokamak vacuum vessel, in-vessel components, and in the components of the neutral beam injector during plasma operations and du...

The ITER International Organization has developed a number of reference Monte Carlo N-Particle (MCNP) models including the tokamak machine C-model, the Tokamak Complex model, and the neutral beam injection (NBI) systems model. The Tokamak Complex model primarily describes building structures beyond the bioshield. Representation of the tokamak and its systems are not included in this model. The Oak Ridge National Laboratory Radiation Transport Group has conducted two ITER neutronic analysis model integrations: (1) integration of the tokamak C-model with the Tokamak Complex model for shutdown dose rate characterization in Port Cell 16 at level B1, and (2) integration of the NBI model with the Tokamak Complex model for estimating the spatial distribution of biological dose rate at levels L1, L2, and L3 of the Tokamak Complex. The integrated models were further extended to include models of system components that are essential to the neutronic analyses. This paper presents the approach and computer to...

The private sector’s recent interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. The measurements are supplemented with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECD NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. This paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.

The Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS) is used to perform dose rate calculations for spent nuclear fuel (SNF) transportation packages based on the actual physical and nuclear characteristics (i.e., assembly design, burnup, initial enrichment, and cooling time) of the as-loaded SNF. Nuclear fuel data, transportation package model templates, and SNF canister loading map information residing within the tool facilitate automated generation of SCALE input files for radiation source term and dose rate calculations. Transportation package specific models developed for UNF-ST&DARDS dose rate analyses are described in detail. UNF-ST&DARDS dose rate analyses were performed for over 400 SNF canisters from 16 sites in their designated transportation casks. For simplicity, representative dose rate calculation results are presented as a function of time (i.e., selected calendar years between 2020 and 2100) for 73 SNF canisters in dry storage a...

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages...

To understand the changing nuclear and mechanical characteristics of spent nuclear fuel (SNF) or used nuclear fuel (UNF) and the different storage, transportation, and disposal systems at various stages within the waste management system, different types of analyses are required. These analyses require the use of assorted tools and numerous types of data. Using the appropriate modeling and simulation (M&S) parameters and selecting from the diversity of analytic tools to conduct SNF analyses can be a tedious, error-prone, and time-consuming undertaking for analysts and reviewers alike. A new, integrated data and analysis system was designed to simplify and automate performance of accurate, efficient evaluations for characterizing the input to the overall U.S. nuclear waste management system—the UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database has been assembled to provide a standard means by which UNF-ST&DARDS can succinctly store and re...

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-STD calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associatedmore » with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

New isotopic depletion capabilities and ENDF/B-VII data libraries have been implemented in the recent release 6.1 of SCALE, a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by Oak Ridge National Laboratory. An assessment of the effect of the new developments on the code performance is the subject of this paper. The analysis is focused on evaluating the code performance in predicting isotopic compositions in spent nuclear fuel by using an extensive, measured isotopic assay database. The analysis results obtained using the latest ENDF/B-VII cross-section data and different resonance processing methods in SCALE are compared to the results of previous validation studies that used ENDF/B-V data. The performance of SCALE depletion capabilities with respect to other computational systems is assessed based on recent published results that were obtained using ENDF/B-VII libraries.

The calculation of fuel isotopic compositions is essential to support design, safety analysis, and licensing of many components of the nuclear fuel cycle—from reactor physics and severe accident an...

The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C{sub k}, approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C{sub k} > 0.95) were attainedmore » at or near the end of a reactor cycle. The C{sub k} values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K{sub eff} sensitivity profiles toward higher energies in the thermal region as compared to the K{sub eff} sensitivity profile of the generic cask. Parameters representing coverage of the application by the CRCs on an energy-dependent, nuclide-reaction specific level (i.e., effectiveness of the CRCs for validating the cross sections as used in the application) were also examined. Based on the CRCs with C{sub k} > 0.8 and an assumed relative standard deviation for uncovered covariance data of 25%, the relative standard deviation of K{sub eff} due to uncovered sensitivity data varies from 0.79% to 0.95% for cask burnups ranging from 10 to 60 GWd/MTU. As expected, this uncertainty in K{sub eff} is largely dominated by noncoverage of sensitivities from major actinides and hydrogen. The contributions from fission products and minor actinides are very small and comparable to statistical uncertainties in K{sub eff} results. These results (again, assuming a 25% uncertainty for uncovered covariance data) indicate that there could be approximately 1% uncertainty in the calculated application K{sub eff} due to incomplete neutronic testing (validation) of the software by the CRCs. However, this conclusion also assumes all other uncertainties in the complex CRC configurations (e.g., isotopic compositions of burned fuel, operation history, data) are well known. Thus, an evaluation of the uncertainties in the CRC configurations is needed prior to the use of CRCs for code validation (i.e., quantifying code bias and bias uncertainty).« less

The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister-based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE’s Office of Nuclear Energy Used Fuel Disposition Campaign's Deep Borehole Field Test. Groundwork for Universal Canister System Development September 2015 ii Wastes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system. Future work includes collaboration with the Hanford Site to move the cesium and strontium capsules into dry storage, collaboration with the Deep Borehole Field Test to develop surface handling and emplacement techniques and to develop the waste package design requirements, developing universal canister system design options and concepts of operations, and developing system analysis tools. Areas in which further research and development are needed include material properties and structural integrity, in-package sorbents and fillers, waste form tolerance to heat and postweld stress relief, waste package impact limiters, sensors, cesium mobility under downhole conditions, and the impact of high pressure and high temperature environment on seals design. September 2015 Groundwork for Universal Canister System Development

An evaluation of surface dose rates calculated by MCNP with three different source geometric representations for a waste package containing 21 pressurized water reactor (PWR) spent fuel assemblies is provided. The three geometric representations for the source region consist of a radial homogenization inside the waste package cavity, a homogenization of the assembly contents within the transverse dimensions of the assembly, and a detailed geometric model. Conservative dose rates are predicted by the source homogenized inside the cavity, while the detailed and the homogenized assembly models give comparative results.

Sweet sorghum is a great energy crop that shows the benefits to ecosystems, energy and economics, being a valuable source of energy of the category 1st, 2nd and 3rd. bioethanol generation. Purpose of the paper is to study the chemical composition of sweet sorghum stalks depleted in carbohydrates with applications in obtaining ethanol. It shows appreciable compositional values of free sugars, starch, cellulose, hemicellulose and lignin. All these components can be easily made available as fermentable carbohydrates leading to the production of products with high economic value (bioethanol).

Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensingmore » Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with predicted spent fuel compositions (i.e., determine the penalty in reactivity due to isotopic composition bias and uncertainty) for use in disposal criticality analysis employing burnup credit. The method used in this calculation to propagate the isotopic bias and bias-uncertainty values to k{sub eff} is the Monte Carlo uncertainty sampling method. The development of this report is consistent with 'Test Plan for: Isotopic Validation for Postclosure Criticality of Commercial Spent Nuclear Fuel'. This calculation report has been developed in support of burnup credit activities for the proposed repository at Yucca Mountain, Nevada, and provides a methodology that can be applied to other criticality safety applications employing burnup credit.« less

The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this reportmore » is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.« less

The spent nuclear fuel (SNF) aging system at the proposed Yucca Mountain repository will provide site-specific casks and aging pads for thermal management of commercial SNF with a heat rate in excess of the waste package thermal output limit. An aging pad can accommodate 1,000 MTHM of SNF, containing a total of 100 aging casks with a horizontal module of 20 casks, and 80 vertical site-specific casks arranged in a 2 x 40 array. The proposed aging system will provide five aging areas in two separate locations. The first location will contain a single pad designated as Aging Area 17A (1,000 MTHM capacity). The second location will contain Aging Areas 17B through 17E (20,000 MTHM total capacity), each consisting of five aging pads arranged in a compact rectangular configuration. This paper presents calculated dose rates as a function of distance from Aging Areas 17A and 17B through 17E. In addition, the paper evaluates the effect of design parameter variations on dose rates with focus on spacing between casks and spacing between pads in Aging Areas 17B through 17E.

This paper presents results of criticality calculations performed for three sets of benchmark problems conducted with four computer code systems. The problems presented in this paper include both experimental and computational benchmarks. These calculations were performed for the Amarillo Resource Center for Plutonium (ANRCP) in conjunction with the Oak Ridge National Laboratory (ORNL). Each of the following sections contains an abbreviated problem statement, a brief description of the calculational methodology, and primary results. MCNP-4A with ENDF/B-VI is used for the Monte Carlo calculations. WIMSD4-m/TWODANT, WIMSD4-m/DIF3D, and CASMO-3 are used for different parts of the deterministic calculations. Due to space constraints, the descriptions given below are quite brief. Please contact the authors if further information is desired.

The Saxton critical experiments, which used mixed-oxide (MOX) fuel of 6.6 wt% PuO 2 in natural UO 2 and UO 2 fuel of 5.74 wt% 235 U, are analyzed with MCNP-4B and continuous-energy cross-section libraries ENDF/ B-V and ENDF/B-VI. An excellent agreement of calculated and experimental effective multiplication factors for the entire set of 1.3208-cm MOX lattices and 1.4224-cm MOX and UO 2 lattices was obtained. The analysis of criticality calculations for the five different lattice pitches show a bias with lattice pitch, which led to an increase of ∼0.8% when doubling the lattice pitch. Good agreement between calculated and measured data was obtained for some of the relative power distribution experiments for MOX single-region cores and MOX/UO 2 multiregion cores; however, for others the agreement was less satisfactory. No significant difference in the results for relative power with the two libraries was observed.