Spent Nuclear Fuel Assemblies Research Articles

On the use of neutron flux gradient with ANNs for the detection of diverted spent nuclear fuel

One of the main tasks in nuclear safeguards is regular inspections of Spent Nuclear Fuel (SNF) assemblies to detect possible diversions of special nuclear material such as 235U and 239Pu. In these inspections, characteristic signatures of SNF such as emissions of neutrons and gamma rays from the radioactive decay, are measured and their consistency with the declared assemblies is verified to ensure that no fuel pins have been removed. Research in this field is focused on both the development of detection equipment and methods for the analysis of the acquired measurement data. In this paper, the use of the neutron flux gradient, which is not considered in regular SNF verification, is investigated in combination with the scalar neutron flux as input to artificial neural network models for the quantification of fuel pins in SNF assemblies. The training and testing of these ANN models rely on a synthetic dataset that is generated from Monte Carlo simulations of a typical intact pressurized water reactor assembly and with different patterns of fuel pins replaced by dummy pins. The dataset consists of unique scenarios so that the ANN can be assessed over “unknown” cases that are not part of the learning phase. Results show that the neutron flux gradient is advantageous for a more accurate reconstruction of diversions within SNF assemblies.

Preliminary data analysis of surrogate fuel-loaded road transportation tests under normal conditions of transport

In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

Experimental Verification of Proposed Internal Remote Monitoring System for Spent Nuclear Fuel Dry Cask Storage Applications

It has become a common practice to store sufficiently cooled spent nuclear fuel (SNF) assemblies in interim storage dry casks with passive cooling. These dry casks require nuclear safeguards monitoring because they contain plutonium. Past studies on dry cask modeling and simulations have shown that a remote monitoring system (RMS) situated inside the dry cask could continually monitor and detect the removal of even a single SNF assembly from the cask. This conceptual RMS design was tested by conducting laboratory-scale experiments using small-size 252Cf neutron sources. These small-size sources were surrounded by neutron-reflecting materials in the experiments to mimic the SNF assemblies as a surface neutron source to the fission chamber detectors of the RMS. Experimental and simulation results showed that the removal or diversion of even a single neutron source is detectable within 4 min with a probability of detection greater than 80%.

Design and Analysis of an Internal Remote Monitoring System for Spent Nuclear Fuel Stored in a Dry Cask

Until a long-term solution is finalized, interim storage of sufficiently cooled spent nuclear fuel (SNF) assemblies in dry casks is the predominant practice. Since these dry casks can contain approximately 160 kg of reactor-grade plutonium, they require safeguards monitoring. Results of a simulation study conducted on the design development and analysis of a remote monitoring system (RMS) are presented. The goal of the study was to determine the suitability of this RMS to meet the SNF monitoring objectives. MCNP simulations of a dry cask with all its contents and a set of simulations with one or two removed SNF assemblies were performed to test the detection capabilities of the RMS. The removed assemblies were substituted with dummy assemblies to simulate concealment. The studies showed that the RMS design is suitable to monitor and detect the removal of even a single SNF assembly from the cask.

Simultaneous reconstruction of emission and attenuation in passive gamma emission tomography of spent nuclear fuel

In the context of international nuclear safeguards, the International Atomic Energy Agency (IAEA) has recently approved passive gamma emission tomography (PGET) as a method for inspecting spent nuclear fuel assemblies (SFAs). The PGET instrument is essentially a single photon emission computed tomography (SPECT) system that allows the reconstruction of axial cross-sections of the emission map of an SFA. The fuel material heavily self-attenuates its gamma-ray emissions, so that correctly accounting for the attenuation is a critical factor in producing accurate images. Due to the nature of the inspections, it is desirable to use as little a priori information as possible about the fuel, including the attenuation map, in the reconstruction process. Current reconstruction methods either do not correct for attenuation, assume a uniform attenuation throughout the fuel assembly, or assume an attenuation map based on an initial filtered back-projection reconstruction. We propose a method to simultaneously reconstruct the emission and attenuation maps by formulating the reconstruction as a constrained minimization problem with a least squares data fidelity term and regularization terms. Using simulated data, we show that our approach produces clear reconstructions which allow for a highly reliable classification of spent, missing, and fresh fuel rods.

Applicability of one-body model to predict thermal distribution in spent nuclear fuel pool

An assumption system with a one-body model of spent nuclear fuel (SNF) assemblies was developed to predict the thermal distribution in the spent fuel pool (SFP) in the absence of external cooling system. It was based on the three-dimensional (3D) thermal hydraulic behaviour computed using the computational fluid dynamic (CFD) software, Fluent 18.0. This study aims to determine the applicability of the one-body model (average decay heat) by comparing it with the individuals-body model (variation of decay heat) computation results. The thermal distribution for both models were compared and evaluated, neglecting other effects such as position of the SNF assemblies, decay heat, and axial heat flux profile. It was found that the one-body model was applicable to predict the thermal distribution of the SFP during loss of active cooling system with error less than 1°C.

A numerical study of spent nuclear fuel dry storage systems under extreme impact loading

The structural integrity of concrete dry storage systems under an impact loading caused by a simulated commercial airplane engine crash is evaluated in this study. A finite element model of a generic vertical dry storage system including the concrete overpack, multi-purpose canister, spent nuclear fuel (SNF) basket, and SNF assemblies was developed using a commercial finite element analysis software. The strain rate effect was considered for the steel parts of the cask via a rate dependent and nonlinear constitutive model. The modeling approach was first validated using experimental data available in literature on projectile impact of concrete and steel/concrete composite (sandwich) panels. Then, a comprehensive comparison in terms of strains, damage and deformations was performed for two configurations of reinforced and sandwich cask overpacks commonly used in the industry. Critical impact height was identified by an examination of the strains and accelerations in the SNF assemblies. Furthermore, the effect of boundary conditions for the two cases of free-standing and anchored configuration was investigated. The maximum deformation and stress as well as the peak acceleration were identified in the SNF basket to assess safety of the SNF assemblies. The minimum impact velocity causing tip-over from aircraft engine impact was determined.

Innovative technologies for spent fuel safe management at Ignalina channel-type reactors

Abstract In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for “Channelized Large Power Reactor”-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

Heat Transfer Modeling of Spent Nuclear Fuel Using Uncertainty Quantification and Polynomial Chaos Expansion

A novel method that incorporates uncertainty quantification (UQ) into numerical simulations of heat transfer for a 9 × 9 square array of spent nuclear fuel (SNF) assemblies in a boiling water reactor (BWR) is presented in this paper. The results predict the maximum mean temperature at the center of the 9 × 9 BWR fuel assembly to be 462 K using a range of fuel burn-up power. Current related modeling techniques used to predict the heat transfer and the maximum temperature inside SNF assemblies rely on commercial codes and address the uncertainty in the input parameters by running separate simulations for different input parameters. The utility of leveraging polynomial chaos expansion (PCE) to develop a surrogate model that permits the efficient evaluation of the distribution of temperature and heat transfer while accounting for all uncertain input parameters to the model is explored and validated for a complex case of heat transfer that could be substituted with other problems of intricacy. UQ computational methods generated results that are encompassing continuous ranges of variable parameters that also served to conduct sensitivity analysis on heat transfer simulations of SNF assemblies with respect to physically relevant parameters. A two-dimensional (2D) model is used to describe the physical processes within the fuel assembly, and a second-order PCE is used to characterize the dependence of center temperature on ten input parameters.

As-Loaded Criticality Margin Assessment of Dual-Purpose Canisters Using UNF-ST&DARDS

A novel assessment has been completed to determine the previously unquantified and uncredited criticality margin available in as-loaded commercial spent nuclear fuel (SNF) canisters. This assessment was performed as part of a broader effort to assess issues and uncertainties with storage, subsequent transportation, and final disposal of SNF canister systems. Detailed analyses crediting the burnup, initial enrichment, and postirradiation cooling time of actual SNF inventory were performed for 554 SNF canisters stored at 23 commercial reactor sites to determine realistic criticality safety margins. These detailed analyses were automated by the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS), a comprehensive, integrated data and analysis tool. Calculated, uncredited criticality margins (Δkeff) with respect to the safety analysis results range from 0 to almost 0.30 Δkeff for normal storage and transportation cases. Calculated eigenvalues (keff) range from 0.72 to 1.11 assuming a degraded neutron absorber disposal condition, and they range from 0.94 to 1.20 assuming a degraded basket disposal condition. Calculations with NaCl present in the moderator (which is possible for certain disposal geologies) were used to demonstrate the possibility for subcriticality for degraded cases with a keff above 0.98 with freshwater. The methods used to calculate keff for the canisters analyzed in this work are discussed in detail.The results demonstrate that, for the majority of canisters analyzed here, significant uncredited safety margin is available that could be used to compensate for uncertainties in the SNF assembly and canister internal components. These uncertainties are associated with long-term storage and subsequent transportation and disposal. Results also suggest that the inherent margins associated with how canisters are loaded could support future changes in licensing SNF storage and transportation systems to directly or indirectly credit the margins associated with actual SNF characteristics. Ongoing research continues to gather additional data to quantify uncredited safety margins for SNF canisters loaded at other nuclear reactor sites and to explore potential methods for applying this uncredited margin.

Heat transfer enhancement for spent nuclear fuel assembly disposal packages using metallic void fillers: A prevention technique for solidification shrinkage-induced interfacial gaps

This study considers replacing the externally accessible void spaces inside a disposal package containing a spent nuclear fuel assembly (SNFA) with high heat conducting metal to increase the effective thermal conductivity of the package and simplify the heat transfer mechanism inside the package by reducing it to a conduction dominant problem. The focus of the study is on preventing the gaps adjacent to the walls of the package components, produced by solidification shrinkage of poured liquid metal. We approached the problem by providing a temporary coating layer on the components to avoid direct build-up of thick metal oxides on their surface to promote metallic bonding at the interfaces under a non-inert environment. Laboratory scale experiments without SNFA were performed with Zn coated low carbon steel canisters and Zamak-3 void filler under two different filling temperature conditions – below and above the melting point of Zn (designated BMP and AMP respectively). Gap formation was successfully prevented in both cases while we confirmed an open gap in a control experiment, which used an uncoated canister. Minor growth of Al-Fe intermetallic phases was observed at the canister/filler interface of the sample produced under the BMP condition while their growth was significant and showed irregularly distributed morphology in the sample produced under the AMP condition, which has a potential to mitigate excessive residual stresses caused by shrinkage prevention. A procedure for the full-scale application was specified based on the results.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. Dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.To further evaluate the intensity of contact interaction induced by the local contacts’ impact loading at the spacer grid, detailed models of the actual spring and dimples of the spacer grids were created. The impacts between the fuel rod and springs and dimples were simulated with a 20g transient shock load. The associated contact interaction intensities, in terms of reaction forces, were estimated from the finite element analyses (FEA) results. The bending moment estimated from the resultant stress on the clad under 20g transient shock can be used to define the loading in cyclic integrated reversible-bending fatigue tester (CIRFT) vibration testing for the equivalent condition. To estimate the damage potential of the transient shock to the SNF vibration lifetime, drop tests were performed on the CIRFT specimens. FEA was used to investigate the contact reaction at CIRFT test samples during impact loading induced by drop tests, and the result was compared with that from a 20g acceleration transient shock load.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

A novel assessment has been completed 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 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 and Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed, 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 associated 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.

Overview of the Lithuanian programme for disposal of RBMK-1500 spent nuclear fuel

AbstractIn Lithuania all the spent nuclear fuel (SNF) came from operation of the Ignalina nuclear power plant with two reactors of RBMK type (RBMK is a Russian acronym for 'Channelized Large Power Reactor' which is a water-cooled graphite-moderated reactor: RBMK-1500). Approximately 22,000 SNF assemblies are due for geological disposal in Lithuania. Currently it is envisaged that SNF will be stored in dry interim storage facilities (new and existing) for at least 50 y prior to possible deep geological disposal.The decision on the final SNF management option (disposal in a national repository, disposal in regional repositories, etc.) has not yet been made but some investigations of the possibilities to dispose of the SNF in Lithuania have been initiated. With the support of Swedish experts, analysis of possible geological formations for SNF disposal was performed and the existence of potentially suitable formations agreed. The geological formations prioritized as prospective include the crystalline rocks in southern Lithuania and two clayey formations: the Lower Triassic clay formation and the Lower Cambrian Baltic Group clay formation, with priority given to the Lower Triassic clay formation.This paper presents the main aspects of the research and other activities undertaken over the past decade in the field of SNF disposal: international cooperation; current status and plans for the Lithuanian national program; further investigations required; and competence developments.

Uncertainty and sensitivity analysis of radionuclide migration through the engineered barriers of deep geological repository: Case of RBMK-1500 SNF

Like other nuclear countries, Lithuania is facing the problem of spent nuclear fuel (SNF) disposal from RBMK11“RBMK” is a Russian acronym for “Channelized Large Power Reactor”.-1500 type reactors at the Ignalina NPP (~22,000 of the SNF assemblies). The application of probabilistic and deterministic methods for uncertainty assessment and important parameter ranking is of high importance in order to build confidence in the assessment results on radionuclide migration and indicates directions for further uncertainty reduction.This study is devoted to the uncertainty and sensitivity analysis of radionuclide migration through the engineered barriers of a conceptual geological repository for RBMK-1500 SNF disposal. The analysis was performed using computer codes AMBER (United Kingdom) and MATLAB (Sweden). The analysis showed that the uncertainty of defect enlargement time had no significant impact on 129I and 226Ra peak flux from the engineered barriers yet it was an important parameter for time dependent mean flux of 129I. Sensitivity analysis revealed that in the case of 129I peak flux, further investigations should be oriented to more precise determination of diffusion coefficient in bentonite, instant release fraction and SNF dissolution rate. In the case of 226Ra, efforts to reduce the uncertainty of SNF dissolution rate, sorption coefficient in bentonite and equivalent flow rate should be made.

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000MWth output and tritium self-sufficiency (TBR>1.15), while the required maximum fusion powers are lower than 500MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the 134Cs/137Cs, Pu, and U isotope ratio measurements.

Uncertainty Quantification for New Approaches to Spent Fuel Assay

Estimating plutonium (Pu) mass in spent nuclear fuel assemblies (SFAs) helps inspectors ensure that no Pu is diverted. Therefore, nondestructive assay (NDA) methods are being developed to assay Pu mass in SFAs. Uncertainty quantification is an important task in most assay methods, and particularly for SFA assay. A computer model (MCNPX) is being used to predict isotope masses and the spatial distribution of masses in virtual SFAs for 64 combinations of initial fuel enrichment (IE), fuel utilization [burnup (BU)], and cooling time (CT) values. Additional MCNPX modeling for the same 64 virtual SFAs provided the expected detector responses (DRs) for several NDA techniques such as the passive neutron albedo reactivity method and the 252Cf interrogation with prompt neutrons method.A previous paper describes one uncertainty quantification approach involving Monte Carlo (MC) simulation using individually any of six new NDA options together with IE, BU, and CT. This paper provides an interpretation of the MC approach that is suited for a numerical Bayesian alternative, separately assesses the impact of MCNPX interpolation error, and compares several options to use subsets of IE, BU, CT, and one DR.

Heuristic Methods for Assigning Spent Nuclear Fuel Assemblies to Canisters for Final Disposal

The disposal of spent fuel assemblies (SFAs) by companies currently producing nuclear power in Finland is the responsibility of a company named Posiva Oy. Posiva Oy has decided to use the KBS-3 (Swedish abbreviation for nuclear fuel safety; version 3) concept. In KBS-3, SFAs are placed in metal canisters, which are themselves deposited deep into crystalline rock. The disposal process in Finland will last many decades. To efficiently assign SFAs to canisters, in this paper we study the minimax canister formation problem. In this problem, we assume we are given two sets of data: (a) a schedule specifying the number of disposal canisters per year and (b) the decay heat of each SFA for every disposal year. The goal in the problem is to assign SFAs to canisters so that the largest canister heat load is minimized. The minimax canister formation problem is a variant of a well-known optimization problem: makespan minimization on unrelated parallel machines. We developed heuristic methods for solving the minimax canister formation problem. Using our methods and predicted SFA amounts and properties for Finland, we obtained high-quality solutions in numerous test cases. We also investigated how the uncertainty in SFA burnups affects the canister heat loads.