Spent Fuel Assemblies Research Articles

Neutronic Characterization for a Pressurized Water Reactor Spent Fuel Assembly

The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessing or final disposition requires information about its isotopic composition, decay heat, and other thermomechanical properties. The spent nuclear fuel assembly of a typical advanced pressurized water reactor, AP-1000, was characterized using the Monte Carlo MCNPX code and SCALE/ORIGEN code. The simulation of operational history started from the operation of the first fresh core for an average fuel assembly with certain physical isotopic parameters until 25 GWd/tonne U discharge burnup. The analysis considered the calculations of the radionuclide inventories, activity, neutron emission spectrum, gamma-ray emission spectrum, and decay power after 700 effective full power days and for post different time ranges until a 1 million–year cooling period. The comparison of some results of the two codes showed small differences due to the consideration of the continuous-energy variation for neutrons in the MCNPX code and the discrete energy assumption in the SCALE/ORIGEN code.

Simulation of Spent Fuel Assembly Transport Vertical Drop Based on LS-DYNA

The finite element technology is used to carry out the drop simulation and safety assessment of the spent fuel assembly and its transport cask. In view of the problem of how to reasonably simplify the finite element model, a more reasonable and accurate multistage simplified model is proposed for the spent fuel assemblies packed densely in the transport cask, which needs to be analyzed synchronously when dropping. First, the “key” spent fuel assembly is retained, and the “unimportant” assembly is simplified as a cylindrical entity. Based on this, the structure will continue to be simplified and a 1/4 near-earth model containing all characteristics will be established. Then, the LS-DYNA software is used for explicit dynamic analysis of vertical drop conditions. It is concluded that the failure risk of the spacers is the greatest. Therefore, the purpose of this paper is to propose a multistage simplified method for spent fuel assembly and its transport cask, which can provide a reference for the structural integrity analysis of the spent fuel assembly transport drop conditions.

Analysis of loss of cooling accidents in the spent fuel pools using system thermal hydraulics and computational fluid dynamic codes

The interest in the safety of on-site storage of Spent Fuel Assemblies (SFAs) has been increased after the accident on 11 March 2011, in Fukushima Daiichi Nuclear Power Plant. It is important to analyse the behaviour of not only integral effects in spent fuel pools, but also the local effects during heat removal from the SFAs. Usually, for the integral analysis the one-dimensional system computer codes were used. These codes allow to analyse whole pools and cover long time, with a relatively low cost of computational resources. The Computational Fluid Dynamic (CFD) codes could perform very detailed analysis, but this requires large computational resources. The strong side of CFD codes is the possibility to assess local effects which could occur.In the current work the analysis of heat transfer from the SFAs, placed in special baskets was performed using ANSYS FLUENT 17.2 and RELAP5 codes. The models of single basket with SFAs were developed using both codes. The comparison of calculations allowed to evaluate capabilities of both codes. It is shown, that using one-dimensional system thermal hydraulics code developed model allows to describe coolant mixing inside basket sections and received results fits well with CFD code calculation results. Based on gained experience, the loss of cooling accident in the spent fuel pools of Ignalina NPP was analysed using the model, developed by employing RELAP5 (one-dimensional system thermal hydraulics) code. The calculation results were used for safety evaluation of Ignalina NPP facilities.

DISTRIBUTION OF RADIOACTIVE ELEMENTS DURING SLAG REMELTING OF STRUCTURAL MATERIALS OF VVER FUEL ELEMENT SHELLS

At the radiochemical plant of FSUE “PO “Mayak” in the process of processing spent nuclear fuel (SNF), about 170 tons/year of metal radioactive waste (MRW) are formed, mainly represented by fuel element shells structural materials (SM) containing residual amounts of radionuclides after fuel dissolution, and spent fuel assemblies (SFAS) sent to storage. It is possible to achieve compaction and decontamination of MRW by a method based on slag remelting in an induction furnace with a cold crucible. In order to establish the distribution of actinides and fission products (FP), thermodynamic slag remelting process modeling of the VVER-1000 reactor plant SM fuel rods shells and SFAS, experimental data obtained verification were carried out. The most probable distribution of actinides and FP by melting products is shown. Most of the curium and americium are concentrated in the metal – 99 and 94 wt %, respectively. The maximum extraction of uranium into the slag phase in the form of UO2 dioxide is up to 40 wt %. The distribution of plutonium in the temperature range of 1500–2000°C over the slag (in the form of PuO and PuO1.61 oxides) and metal phases occurs in almost equal proportions. It has been established that up to 99.78 wt % cesium, the content of europium and americium is 0.05 and 0.17 wt %, respectively.

Radiation shielding analysis of the HI-STORM FW storage cask

This paper presents radiation shielding analysis of the HI-STORM FW storage cask designed by the Holtec International, Inc. for the Nuclear Power Plant Krsko Spent Fuel Dry Storage (SFDS) project. The first part of the analysis is addressing the problem of a particle source generation. For each spent fuel assembly (SFA) neutron and gamma source intensities and spectra were calculated using ORIGEN-S module of the SCALE 6.2.3 code package. The second part of the analysis is considering neutron and gamma dose rates calculation, for which hybrid shielding methodology based on Monte Carlo (MC) code MCNP6.2 and ADVANTG3.0.3 code was used. The dose rates for the reference case included the contributions of fuel neutron, fuel gamma, secondary gammas from (n, γ) reactions, as well as gammas from activated SFA hardware components due to Co-60 presence. Expectedly, the results have shown that the dose rates around the cask have nonuniform behavior due to a specific design of the cask. The heavy concrete body of the cask presents a good attenuator for all types of radiation, however, a considerable streaming through the top and bottom ventilation openings was noticed. It was also observed that an important role was played by the isotope Co-60 which emits two lines of high energy deep penetrating gammas. The convergence of dose rates and associated relative uncertainties during MC calculations have shown that significantly more histories were required for gamma calculations compared to neutron calculations. To investigate the impact of different SFA content in the casks on the dose rates, shielding analysis was made for 16 casks that will be loaded in the first campaign. For comparison, the calculations of a single HI-STORM FW cask were additionally performed using MAVRIC module of the SCALE 6.2.3 and the obtained differences in external dose rates of 20–30% are expected. The calculation cases showed the differences in casks doses due to different casks loading within loading plan criteria, reduction provided by heavy concrete in cask body (more important for gammas), importance of careful design of lower and upper casks openings to keep increase of dose rates, compared to closed configuration, within one order of magnitude, and compromise needed when describing details of intensity and spectrum of SFA sources. The increased calculation mesh density was used to show that selected discretization is appropriate, and the limited influence of storage pad thickness was demonstrated by performing calculation with thick concrete slab corresponding to Dry Storage Building (DSB) floor foundation thickness.

Preliminary study of artificial intelligence-based fuel-rod pattern analysis of low-quality tomographic image of fuel assembly

Single-photon emission computed tomography is one of the reliable pin-by-pin verification techniques for spent-fuel assemblies. One of the challenges with this technique is to increase the total fuel assembly verification speed while maintaining high verification accuracy. The aim of the present study, therefore, was to develop an artificial intelligence (AI) algorithm-based tomographic image analysis technique for partial-defect verification of fuel assemblies. With the Monte Carlo (MC) simulation technique, a tomographic image dataset consisting of 511 fuel-rod patterns of a 3 × 3 fuel assembly was generated, and with these images, the VGG16, GoogLeNet, and ResNet models were trained. According to an evaluation of these models for different training dataset sizes, the ResNet model showed 100% pattern estimation accuracy. And, based on the different tomographic image qualities, all of the models showed almost 100% pattern estimation accuracy, even for low-quality images with unrecognizable fuel patterns. This study verified that an AI model can be effectively employed for accurate and fast partial-defect verification of fuel assemblies.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Abstract In this paper, shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW (Holtec International Transfer Cask Variable Weight) loaded with spent fuel assemblies (SFA) from nuclear power plant (NPP) Krsko spent fuel dry storage (SFDS) campaign one. The HI-TRAC VW is a multilayered cylindrical vessel designed to accept a multipurpose canister (MPC) during loading, unloading, and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using origen-s module of the scale code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron-induced gammas (n-γ reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, forward weighted consistent adjoint driven importance sampling (FW-CADIS) variance reduction scheme, as implemented in advantg code, was adopted for accelerating final mcnp6.2 calculations. The dose rates around HI-TRAC VW cask were calculated using mcnp6.2 code for all 16 cask loadings belonging to campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization - Holtec International (HI).

Structural integrity investigations of spent nuclear fuel with finite element modeling

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

Estimation of the Duration of Cooling Time of the SFA of the VVER-1200 Reactor Depending on the Type of Transport Container

As the burnup increases, the requirements of nuclear and radiation safety for spent nuclear fuel (SNF) at the subsequent stages of operation grow. The analysis and estimation of the required cooling time of the spent fuel assemblies (SFAs) of the VVER-1200 reactor for transportation in various types of transport containers (TUK-13 and TUK-141O) are performed. The estimates are based on the analysis of residual energy release and gamma-radiation intensity of SNF depending on the cooling time for different burnups. The data on the absorbed dose rate of neutron and gamma radiation from SNF after 4 and 5 yr of cooling are presented.

SAFETY CULTURE IN TECHNOLOGIES FOR HANDLING NUCLEAR WASTE AND NUCLEAR FUEL

The safety culture in the elimination of the nuclear legacy from the defense and civil programs of the former USSR is demonstrated by the example of three different technologies for handling nuclear waste and spent nuclear fuel. These technologies have been implemented by RFNC-VNIITF together with RFNC-VNIIEF, OJSC RI, JSC NIKIET, and the National Nuclear Center of the Republic of Kazakhstan (NCC RK). The export of nuclear fuel from the NCC RK research reactors to Russia is considered. The key aspects of eliminating access to nuclear waste at the former Semipalatinsk Test Site in the interests of the non-proliferation regime are discussed. The set of activities associated with the export of the spent fuel assemblies (SFAs) of reactors AMB-100 and AMB-200 as the nuclear legacy of the first civil nuclear power facilities to the “Mayak” PA is shown. It is noted that compliance with the national laws, regulatory documents, and safety culture has made it possible to successfully perform unique technical and organizational tasks. As a result, the Republic of Kazakhstan has met all the requirements for a non-nuclear state when staging research reactor complexes under the IAEA safeguards; the Russian Federation, the Republic of Kazakhstan, and the United States have jointly eliminated the threat to non-proliferation regime at the former Semipalatinsk Test Site, the State Atomic Energy Corporation “Rosatom” has eliminated the “sluggish accident” at one of the sites of the nuclear legacy of the civil nuclear power engineering.

A new approach to the recycling of spent nuclear fuel in thermal reactors within the REMIX concept

A review of simulated nuclear fuel cycles with mixed uranium-plutonium fuel (REMIX) was carried out. The concept of REMIX fuel is one of the options for closing the nuclear fuel cycle (NFC), which makes it possible to recycle uranium and plutonium in VVER-1000/1200 thermal reactors at a 100% core loading. The authors propose a new approach to the recycling of spent nuclear fuel (SNF) in thermal reactors. The approach implies a simplified fabrication of mixed fuel when plutonium is used in high concentration together with enriched natural uranium, while reprocessed uranium is supposed to be enriched and used separately. The share of standard enriched natural uranium fuel in this nuclear fuel cycle is more than 50%, the share of mixed natU+Pu fuel is 25%, the rest is fuel obtained from enriched reprocessed uranium. It is emphasized that the new approach has the maximum economic prospect and makes it possible to organize the fabrication of this fuel and nuclear material cross-cycling at the facilities available in the Russian Federation in the short term. This NFC option eliminates the accumulation of SNF in the form of spent fuel assemblies (SFA). SNF is always reprocessed with the aim of further using the primary reprocessed uranium and plutonium. Non-recyclable in thermal reactors, burnt, reprocessed uranium, the energy potential of which is comparable to natural uranium, as well as secondary plutonium intended for further use in fast reactors, are sent as reprocessing by-products to the storage area.

Prevention of Damage to Spent Nuclear Fuel during Handling Operations

According to regulatory requirements, when carrying out handling operations with spent nuclear fuel (SNF), prevention of damage to the spent fuel assemblies (SFA) and especially fuel elements shall be ensured.
 For this purpose, it is necessary to exclude the risk of SFA falling, SFA uncontrolled displacements, prevent mechanical influences on SFA, at which their damage is possible.
 Special requirements for handling equipment (in particular, cranes) to exclude these dangerous events, the requirements for equipment strength, resistance to external impacts, reliability, equipment design solutions, manufacturing quality are analyzed in this work. The requirements of Ukrainian and U.S. regulatory documents also are considered.
 The implementation of these requirements is considered on the example of handling equipment, in particular, spent nuclear fuel storage facilities.
 This issue is important in view of creation of new SNF storage facilities in Ukraine. These facilities include the storage facility (SFSF) for SNF from water moderated power reactors (WWER): a Сentralized SFSF for storing SNF of Rivne, Khmelnitsky and South-Ukraine Nuclear Power Plants (СSFSF), and SFSF for SNF from high-power channel reactors (RBMK): a dry type SFSF at Chornobyl nuclear power plant (ISF-2). After commissioning of these storage facilities, all spent nuclear fuel from Ukrainian nuclear power plants will be placed for long-term “dry” storage. The safety of handling operations with SNF during its preparation for long-term storage is an important factor.

Design of Poolside Inspection System for Spent Fuel Assemblies Based on Photogrammetry

Design of Poolside Inspection System for Spent Fuel Assemblies Based on Photogrammetry

Моделирование естественной конвекции у поверхности отработанных тепловыделяющих сборок в бассейне выдержки АЭС

Моделирование естественной конвекции у поверхности отработанных тепловыделяющих сборок в бассейне выдержки АЭС

Conceptual design of neutron measurement system for input accountancy in pyroprocessing

Abstract One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different 235U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

Investigation of the thermal state of an asynchronous motor with an asymmetric magnetic circuit

Modern technology for the processing of spent fuel assemblies (SFA) and radioactive waste (RW) is utilized in radiation protection cells. Under the conditions of exposure to high radiation fields, there is an intense destruction of various assemblies of actuators made of organic materials, including insulation of motor windings, which leads to frequent equipment failures requiring repair or replacement. Within the limits of the existing need, a project is being developed, with the purpose of creation of a radiation-resistant asynchronous electric motor with ceramic insulation for operation in. The proposed technical solutions make possible to increase the service life of the engine operating under the influence of radiation fields in radiation protection chambers by 4-5 times. Operation under an extreme condition requires an assessment of the engine itself. Considering a non-standard cooling system is of crucial importance for such projects.

Analysis of Depletion and Inventory of the Fuel for the MNSR Research Reactor Using the Deterministic Code DRAGON5

The Syrian miniature neutron source reactor (MNSR), a 30-kW, 90.0% highly enriched uranium fueled (U-Al) MNSR-type reactor has gone critical. Under operating conditions of 2 h per day for 5 days a week at a peak thermal neutron flux of 1.0 × 1012 n/cm2·s, the estimated core life is 10 years. After the fuel is depleted, the full spent-fuel assembly will be replaced with new low-enriched uranium. This study presents the results of a multigroup fuel burnup and depletion analysis of the MNSR fuel lattice using the DRAGON5 transport lattice code. Furthermore, infinite multiplication factor k∞ and several two-group macroscopic parameters, including scattering cross section, fission cross section, total cross section, and diffusion coefficient, and the transport mean free path have been studied. In addition to this, fuel isotopic composition dependency on burnup was calculated as a part of this study. The results contained in this study can be used as a microscopic database for performing criticality safety analysis and shielding computations for the design of a spent-fuel storage cask for the MNSR reactor core.

Experimental and analytical study on heat dissipation mechanisms during spent fuel unloading in sodium fast reactor

In a sodium-cooled fast reactor spent fuel assemblies (SFAs) are unloaded from the reactor sodium pool after certain period of temporary storage. During unloading a spent fuel assembly is transferred through argon atmosphere where heat dissipation from SFA to argon is of great concern to avoid cladding damage which is strictly prohibited. With a 37-element rod array the steady-state experiments are carried out to investigate heat transfer under different environment temperatures and power levels. The rod array is installed vertically in a pressure vessel filled with argon. It is found that the circulation flow rate through the rod array driven by buoyancy is negligible. Instead of convection, conduction plays an important role in the radial heat transfer in the rod bundle. The Manteufel-Todreas model is proved to give satisfactory prediction for in-wrapper heat transfer. Natural convection on the outer wrapper surface can be reasonably predicted with the Yang model.