Organisation For Economic Co-operation And Development/Nuclear Energy Agency Research Articles

Toward a Better Understanding of Reflood Thermal Hydraulics: A Summary of the OECD/NEA RBHT Project

Reflood thermal hydraulics remains a difficult and complex subject, and understanding the physical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer (RBHT) project was designed to provide unique experimental data for code assessment and model development. Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility. The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were provided to participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thus producing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcooling ranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the test matrix. This paper describes the project and presents a summary of major experimental and analytical findings.

Evaluation of Molten Corium Spreading and Sedimentation Behaviors Within Primary Containment Vessel in Unit 3 of Fukushima Daiichi Nuclear Power Plant Toward the Best Prediction of Fuel Debris Distribution

Accomplishing the retrieval of fuel debris from Fukushima Daiichi Nuclear Power Plant (1F) Unit 3 (1F3) requires an understanding of its distribution. In this study, we performed real-scale corium spreading and sedimentation behavior analyses using Lagrangian moving particle hydrodynamics and large eddy simulation methods. These methods allowed us to calculate the spreading of corium with various shear viscosities under water conditions and to propose the best estimation for the fuel debris distribution in 1F3. To minimize uncertainties arising from unknown boundary conditions, we investigated relevant parameters through literature review. Our analyses showed that highly viscous corium tends to pile up within the pedestal region under strong convective vapor and boiling heat transfer, while low-viscosity corium spreads to the outside of the pedestal regions regardless of cooling efficiency. We identified three cooling modes based on initial shear viscosity and cooling efficiency and predicted the fuel debris distribution in 1F3 by comparing our results to those of the Tokyo Electric Power Company (TEPCO) and Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. The distribution estimation of highly viscous corium derived from oxidic corium is consistent with the three-dimensional reconstructed image by TEPCO and the calculated results by the OECD/NEA BSAF project.

Propagation of Uncertainties due to Nuclear Data for the LCT086 Reactor System

An Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) benchmark has been established over recent years to satisfy an increasing demand from the nuclear community for best-estimate predictions accompanied by uncertainty and sensitivity analyses. The main objectives of the OECD/NEA benchmark activity are to determine uncertainties in modeling for reactor systems using best-estimate methods under steady-state and transient conditions and to quantify the impact of these uncertainties for each type of calculation in multiphysics analyses. In terms of light water reactor analyses, an international uncertainty analysis, “Benchmarks for Uncertainty Analysis in Modelling (UAM) for the Design, Operation and Safety Analysis of LWRs” is currently in progress, being coordinated by the OECD/NEA. In the neutronic phases of the benchmark, the uncertainty due to nuclear data is being studied for various LWR types. The LCT086 benchmark, which is a VVER-type reactor criticality benchmark experiment, has been identified to form part of the validation matrix for the uncertainty methodology development. Resulting from this, the main focus of the present work is to propagate the uncertainty due to the nuclear data for two cases (LCT086/Case1 and LCT086/Case3) presented in the LCT086 benchmark. Both fuel assembly and core models were used for the analysis. The code package used to perform the calculations was SCALE 6.2.1. In particular, the function modules KENO-VI, SAMPLER, and TSUNAMI-3D of SCALE 6.2.1 were used. MCNP 6.1 was also used for continuous-energy criticality calculations. In addition to propagating the uncertainty due to the nuclear data, the uncertainty due to selected input parameters as bounded by the manufacturing tolerances were also propagated so that the modeling methods employed could be verified against those reported in the LCT086 benchmark. The results obtained for the nuclear data uncertainty were further compared with nuclear data uncertainty propagation results for the OECD/NEA UAM Kozloduy-6 reactor system. The uncertainty in the multiplication factor k eff is reported in pcm together with the main contributors to the uncertainty from the nuclear data reported in %Δk / k .

Aquatic Chemistry of Pentavalent Plutonium: Determination of the First Hydrolysis Constant

Up to now, no recommendations have been made by the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA) for the first two hydrolysis constants of pentavalent plutonium. We have determined them, as well as those of Np(V), by means of capillary electrophoresis coupled with inductively coupled plasma mass spectrometry (CE-ICP-MS) in 0.1 M NaCl at 25 °C. The hydrolysis constants of Pu(V) were compared to those of Np(V), for which consensual values have been proposed by the OECD-NEA. The first hydrolysis stability constant for Pu(V) extrapolated at zero ionic strength is, as expected, close to that of Np(V) . We have obtained an excellent agreement with the OECD-NEA value for Np(V) []. Based on eight independent values including our own, we propose a new, robust value for the first hydrolysis of Np(V): . The determination of the second hydrolysis constant for Np(V) by CE-ICP-MS deviates from the value adopted by the OECD-NEA . This discrepancy might be explained by the association of a sodium counter cation with the species . A stability constant is proposed at zero ionic strength and 25 °C for this association: . As a result, it is considered that the second hydrolysis constant for Pu(V), , is also a mixed constant, taking into account both the hydrolysis and the association with a sodium cation.

URANIUM AND THORIUM POTENTIAL FOR INDONESIA'S FUTURE ENERGY SECURITY

Uranium and Thorium are natural elements known for their radioactive properties that can be used as a source of nuclear power. Both Uranium and Thorium are new energy sources, and are classified as alternative nuclear fuels, they are known as green nuclear. Uranium is not a rare metal, it is estimated that there are 5.7 million tons of uranium reserves in the world in 2015 in the Red Book of The Organization for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA). As for Thorium, which is a heavy metal content in the earth's crust, it has reserves of 6.3 million tons in the world. Compared to Uranium, Thorium has a more even distribution and one of them is found in Indonesia. The National Nuclear Energy Agency (BATAN) stated that Indonesia has 79,830 tons of uranium reserves and 136,966 tons of Thorium, which are scattered in the Bangka Belitung Islands, Kalimantan Island and Sulawesi Island. The purpose of this paper is to examine the potential of Uranium and Thorium in Indonesia, including the amount of reserves, their distribution and future prospects in order to support the nation's energy security, because in Indonesia, nuclear is used for peaceful purposes, namely as a power plant.

The Inter-Agency Committee on Radiation Safety—30 years of international coordination of radiation protection and safety matters

The Inter-Agency Committee on Radiation Safety (IACRS) was constituted in 1990, as a forum for collaboration and coordination between international bodies with regards to radiation safety. It consists today of representatives of eight intergovernmental member organizations (the European Commision, Food and Agriculture Organization, International Atomic Energy Agency (IAEA), International Labour Office, Organisation for Economic Co-operation and Development/ Nuclear Energy Agency (OECD/NEA), Pan American Health Organization, United Nations Scientific Committee on the Effects of Atomic Radiation and the World Health Organization) and five observer non-governmental organizations (International Commission on Radiological Protection, International Commission on Radiation Units & Measurements, International Electrotechnical Commission, International Radiation Protection Association and International Organization for Standardization). The IACRS provides a platform for interaction between these relevant international bodies to contribute to a common understanding of the scientific basis and legal framework for the application of the system of radiation protection, towards global harmonisation of radiation safety standards. The IACRS played a key role in the development of the International Basic Safety Standards (BSS) in 1996 and in its revision in 2014. Further, an IACRS specific Task Group—chaired by the IAEA—fosters the implementation of the BSS in a consistent and coherent manner in all Member States of the United Nations. The IACRS operates via a standing secretariat jointly provided by the IAEA and OECD/NEA and is chaired by one of its member organizations on a rotating basis for periods of about 18 months. This approach has proved to be effective and was the foundation for ensuring continuity of the work of the committee and at the same time allowing a rotating leadership for all member organizations. Currently, the IACRS is chaired by the WHO. The International Radiation Safety Framework under which the IACRS works is structured around four main areas: (a) science; (b) principles; (c) standards; and (d) practice. This paper presents briefly the mandates, roles and functions of the various international bodies that are relevant to the four above mentioned areas of work, discusses how these bodies coordinate their actions and complement each other to enhance radiation protection and safety worldwide and describes their contribution to the achievement of the Sustainable Development Goals. The paper also provides an overview of the main accomplishments of the IACRS since its inception 30 years ago, and an outlook on key challenges for its future activities.

EURAD School of Radioactive Waste Management

Abstract. The European Joint Programme on Radioactive Waste Management (EURAD – H2020, grant agreement No. 847593) aims to achieve a step-wise change in European collaboration towards safe radioactive waste management (RWM) through the development of a robust and sustained science, technology and knowledge management (KM) programme. EURAD has three KM work packages. The main goals are to (i) preserve generated knowledge, (ii) transfer knowledge to Member States with early-stage RWM programmes, (iii) transfer knowledge between generations, and (iv) disseminate knowledge. The EURAD work package (WP) on “Training & mobility” aids in achieving these goals through its “School of Radioactive Waste Management”. The School of RWM supports competence building in RWM matters. Currently four distinguished initiatives can be highlighted: the organization of training courses, the hosting of webinars, the coordination of a mobility programme, and actions to support the EURAD PhD community. The School of RWM has a dedicated webpage (http://www.euradschool.eu, last access: 30 September 2021) where all information on its activities can be found. A portfolio of basic and specialized training courses was set up containing an up-to-date list of existing training initiatives and new training courses. The latter are based on a gap analysis performed within the framework of the priorities set in the EURAD Roadmap and on the end-user needs. The School of RWM gives attention to the best-fitted training format and scientific state-of-the-art is guaranteed through the lecturing by EURAD and/or external subject experts. Besides training courses, webinars on specific topics that are of relevance to the entire EURAD community are hosted on a regular basis. These so-called Lunch & Learn sessions are short, informal online get-togethers, triggering discussion and debate. Presentations are given by experts from the EURAD community and beyond, such as from the International Atomic Energy Agency (IAEA) or the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) and can deal with all topics that link to RWM, from very specialized scientific matters to more general and overarching themes. The School of RWM also coordinates a mobility programme, allowing its beneficiaries to perform technical visits to infrastructures from EURAD partners or end-users, to undertake internships and to set up exchange programmes between organizations within EURAD. These activities serve as enhanced training as well as collaboration between all members of the EURAD community. These mobility actions can be complementary to an educational or training programme and/or part of a continuous personal development programme. Last but not least, the initiative was taken to support the EURAD PhD community. As future key players in the field of RWM, they will benefit from the possibilities for early networking with their peers and with established RWM experts. With these four initiatives, the School of RWM, and by extension the “Training & Mobility” WP, is an integral part of EURAD's KM programme. This presentation describes how the School of RWM is linked to the EURAD Roadmap, what its goal and long-term objectives are and how it aids in achieving EURAD's KM goals.

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

The objective of this study is to develop a 3D neutron kinetic (NK) and thermal hydraulic (TH) coupled model using the INL developed code RELAP5-3D©/PHISICS to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA).This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated, the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode, an extensively study has been carried out. Some tests with linear change in coolant mass flow rate demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.

Validation of COCOSYS 2.4v4 AIM module on various single effect and integral experiments

Abstract Nuclear power plant containment is the last barrier, as stated in the worldwide accepted in-depth safety principle, between a nuclear reactor and the environment. During various accident scenarios, fission products can exit the fuel and primary circuit, entraining the containment and later the environment. One of the most important fission products exiting the corrupted fuel pins is iodine. Iodine is a strongly bioactive chemical species with a very complex in-containment physical-chemical behavior. The fission product transport inside the containment as well as the release into the environment takes an important role in the preliminary safety analysis report (PSAR). Analyses presented in PSAR are calculated by numerous computational codes. The validity of these results should be checked. The only chance to do so is to validate the computational tool on validation experiments, which represent single effect behavior as well as a complex behavior on integral tests. Validation experiments presented in this paper are from an international Behavior of Iodine Project (BIP) which was conducted under the Organization for Economic Co-operation and Development – Nuclear Energy Agency (OEC-NEA). The single effect tests are illustrated by four representative experiments, namely the G-01 studying the iodine deposition and resuspension on steel coupon in gas phase, the G-04 studying the iodine deposition and resuspension on painted surface in gas phase, the G-06 studying the humidity effect on iodine deposition on painted coupons in gas phase and finally the AECL-2 studying the iodine deposition on painted coupons in water phase. The integral test is represented by the RTF P9T1, where the complex iodine chemistry including the response to pH change is studied. The scope of this paper is to reveal whether the COCOSYS code is capable to bring satisfactory results of a complex in-containment iodine behavior with recommended and basic setup of computational models and parameters.

Evaluation of the PAR Mitigation System in Swiss PWR Containment Using the GOTHIC Code

The installation of passive autocatalytic recombiners (PARs) in the containment of operating nuclear power plants (NPPs) is increasingly based on three-dimensional studies of severe accidents that accurately predict the hydrogen pathways and local accumulation regions in containment and examine the mitigation effects of the PARs on the hydrogen risk. The GOTHIC (Generation Of Thermal-Hydraulic Information for Containments) code is applied in this paper to study the effectiveness of the PARs installed in the Gösgen NPP in Switzerland. A fast release of a mixture of hydrogen and steam from the hot leg during a total station blackout is chosen as the limiting scenario. The PAR modeling approach is qualified simulating two experiments performed in the frame of the OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) THAI (Thermal-hydraulics, Hydrogen, Aerosols and Iodine) project. The results of the plant analyses show that the recombiners cannot prevent the formation of a stratified cloud of hydrogen (10% molar concentration), but they can mitigate the hydrogen accumulation once formed. In the case of the analyzed fast release scenario, which is characterized by increasing loads with large initial flow rate and high hydrogen concentration values, it is shown that, when a large number of recombiners are installed, the global outcome in relation to the combustion risk does not depend on the details of the single PAR behavior. The hydrogen ignition risk can be fully mitigated in a timeframe ranging from 15 to 30 min after the fast release, according to the dependence of the PAR efficiency model on the adopted parameters.

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

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.

Dynamic Monte Carlo transient analysis for the Organization for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) C5G7-TD benchmark

Abstract With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Units 2 and 3: German Contribution to the OECD/NEA BSAF Project, Phase 1

On behalf of the German Federal Ministry of Economics and Technology, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) participated in the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) project titled Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF). Analysis of the severe accidents (SAs) that happened in the Fukushima Daiichi nuclear power plant (NPP) requires well-qualified methods and codes, e.g., ATHLET-CD and COCOSYS developed and applied at GRS. Coupled ATHLET-CD/COCOSYS analyses for the SA progression during the first days for the similar Units 2 and 3 of Fukushima Daiichi have been provided as the German contribution to the OECD/NEA BSAF project, phase 1. ATHLET-CD is a detailed SA code based on the thermal-hydraulic code ATHLET of GRS to simulate the processes in the reactor circuit before and during core degradation. COCOSYS is focused on the simulation of design basis and SA progression in the containment and the surrounding buildings of the NPP.The focus is on selected results of the SA analyses in the boiling water reactors at the Fukushima Daiichi site especially with regard to the conditions in the torus-shaped wetwell (WW) of the primary containment and specific modeling needs. The GRS results obtained in this OECD/NEA BSAF project, phase 1, are encouraging in terms of capturing essential SA signatures like reactor and containment pressure, reactor water level, and WW temperature history for the first days of the accident in the analyzed Units 2 and 3. A detailed plant model was built up especially with a detailed torus nodalization allowing modeling of relevant phenomena like thermal stratification in the torus water pool and consideration of plant-specific details with regard to local water/steam injections into the torus water pool through safety systems and valves. As a result, the calculated accident progression of the best-estimate analyses for both units follows the accident time line quite closely. This is a prerequisite for reasonable core degradation calculations, as the time window available for the onset of core degradation between known points in time when safety injection stops and mobile pump injection into the reactor starts is small. The analyses are useful to identify areas that require further attention, to define information needs to be gained from the decommissioning, and to define further research needs with regard to experiments and code improvement.

Benchmark Study of the Accident at the Fukushima Daiichi NPS: Best-Estimate Case Comparison

The Great East Japan earthquake occurred on March 11, 2011, at 14:46, and the subsequent tsunami led Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (NPS) beyond a design-basis accident. After the accident, the Japanese government and TEPCO compiled a roadmap toward an early resolution to the accident including, among the main activities, the employment and improvement of existing severe accident (SA) computer codes. In the member countries of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA), SA codes were developed after the accident at Three Mile Island Unit 2 and widely employed to assess NPS status in the postulated SA conditions. Therefore, working plans have been set up with the country members of the OECD/NEA to apply existing SA codes to analyze the accidents at the Fukushima Daiichi NPS Units 1, 2, and 3 and support the decommissioning, constituting an international program named Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF).The objectives of the BSAF project are to analyze the accident progression of Fukushima Daiichi NPS, to raise the understanding of SA phenomena, to contribute to the improvement of the methods and models of the SA codes, and to define the status of the distribution of debris in the reactor pressure vessels and primary containment vessels for decommissioning.The present technical paper summarizes the achievements obtained through a comparison of the results, emphasizing the portions of the accident where all the participants reached a common consensus and identifying still open questions where future work should be directed. Consensus exists on the current condition of Unit 1, where a large fraction of the fuel is assumed to have relocated ex-vessel. On the other hand, larger uncertainties exist for Units 2 and 3, where in-vessel and ex-vessel scenarios produce a reasonable prediction of the accident progression.

Past and Future Research at IRSN on Corium Progression and Related Mitigation Strategies in a Severe Accident

Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Sûreté Nucléaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core–concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors.This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions.In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis.This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management.

Continuous-Energy Sensitivity Coefficients in the MORET Code

A continuous-energy sensitivity coefficient calculation to nuclear data capability has been recently developed in Version 5.C.1 of the MORET Monte Carlo code developed at Institut de Radioprotection et de Sûreté nucléaire (IRSN). The method used for implementation is the differential operator method. In this method, the estimation of the fission source derivatives is replaced by an estimation of the adjoint flux. Both methodology and tallies are described in this paper. The preliminary verification is mainly performed using code-to-code comparisons with the SCALE6.1 and MCNP6.1 software packages. Configurations used for verification are the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Uncertainty Analyses for Criticality Safety Assessment (UACSA) Expert Group benchmarks, the Jezebel International Criticality Safety Benchmark Evaluation Project (ICSBEP) benchmark, and a configuration from the Matériaux en Interaction et Réflexion Toutes Epaisseurs (MIRTE) French proprietary experimental program. Results show good agreement among the different codes.

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD/NEA MHTGR-350 Benchmark

The Parallel and Highly Innovative Simulation for INL Code System (PHISICS) has been under development at Idaho National Laboratory since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) (three-dimensional nodal transport core calculations); MRTAU (Multi-Reactor Transmutation Analysis Utility) (depletion and decay heat generation); and modules performing criticality searches, fuel shuffling, and generalized perturbation. Coupling of the PHISICS code suite to the thermal-hydraulic system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort, the first phase of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MHTGR-350 benchmark has now been completed.The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detail in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 benchmark and presents the results of exercises 2 and 3 defined for phase I. Exercise 2 required the modeling of a stand-alone thermal fluids solution at the end of equilibrium cycle for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The RELAP5-3D results of four subcases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two subcases.The main focus of this paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that includes kinetics feedback on individual “block” level and thermal feedbacks on a triangular submesh. The higher fidelity that can be obtained by this block model is illustrated with comparison results on the temperature, power density, and flux distributions. It is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher-fidelity block model and that the additional model development and run-time efforts are worth the gains obtained in the improved spatial temperature and flux distributions.

TRIPOLI-4 Simulation of Core Physics Parameters for Two 3600-MW(thermal) Sodium-Cooled Fast Reactors

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters—Keff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution—were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly–level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.0 data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared.

RELAP5 analyses on the influence of multi-dimensional flow in the core on core cooling during LSTF cold-leg intermediate break LOCA experiments in the OECD/NEA ROSA-2 Project

Test-2 and Test-7 of the second Organization for Economic Co-operation and Development / Nuclear Energy Agency (OECD/NEA) Rig-of-Safety Assessment (ROSA-2) Project were performed with the Large Scale Test Facility (LSTF), which simulated thermal-hydraulic responses during a pressurized water reactor cold-leg intermediate break loss-of-coolant accident (IBLOCA). Test-2 simulated 17% cold-leg break with single failure of an emergency core cooling system (ECCS). The core liquid level decreased to the bottom at loop seal clearing (LSC), causing high cladding temperature excursion. Test-7 simulated 13% cold-leg break with full injection of the ECCS. Compared to Test-2, the cladding surface temperature in Test-7 was quite low due to greater liquid level recovery after the LSC. To well understand the observed phenomena and to improve the best-estimate code predictive capability, RELAP5 post-test analyses were performed. The RELAP5 analyses employed two core models: one is a single-channel core model that simulates the whole core with one channel of a vertical stack of nine equal-height volumes, and the other is a multiple-channel core model that is composed of three channels in which adjacent vertically stacked volumes are horizontally connected with cross-flow junctions. The analyses with the multi-channel core model predicted better than with the single-channel core model for such parameters as core-collapsed liquid level and cladding surface temperature for both Test-2 and Test-7, by more realistically representing multi-dimensional flow in the core. Such a practical method for better representation of multi-dimensional flows turned out to be important to improve the IBLOCA analysis.

Thermodynamic Properties for Arsenic Minerals and Aqueous Species

Quantitative geochemical calculations are not possible without thermodynamic databases and considerable advances in the quantity and quality of these databases have been made since the early days of Lewis and Randall (1923), Latimer (1952), and Rossini et al. (1952). Oelkers et al. (2009) wrote, “ The creation of thermodynamic databases may be one of the greatest advances in the field of geochemistry of the last century .” Thermodynamic data have been used for basic research needs and for a countless variety of applications in hazardous waste management and policy making (Zhu and Anderson 2002; Nordstrom and Archer 2003; Bethke 2008; Oelkers and Schott 2009). The challenge today is to evaluate thermodynamic data for internal consistency, to reach a better consensus of the most reliable properties, to determine the degree of certainty needed for geochemical modeling, and to agree on priorities for further measurements and evaluations. Recent attention has been directed to arsenic (As) thermodynamic data, partly because of the worldwide recognition of arsenic poisoning in more than 70 countries (Nordstrom 2002; Ravenscroft et al. 2009) and the need to interpret As mobility more quantitatively in groundwater and surface-water systems. Unfortunately, not as many useful thermodynamic measurements have been made on reactions involving As compared to other major solutes and trace elements. Grenthe et al. (1992), when reviewing As data for the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) thermodynamic database on U, stated “ Although needed, a complete reanalysis of the chemical thermodynamic data for arsenic species is not within the scope of the current review .” Although a complete reanalysis is not feasible at this time, this chapter reviews As thermodynamic data with a focus on internal consistency and the quality of the original measurements, updates with new data, …