Central Interim Storage Facility Research Articles

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

Estimating gamma and neutron radiation fluxes around BWR quivers for nuclear safeguards verification purposes

Non-destructive assay (NDA) methods are at the core of nuclear safeguards verification of spent nuclear fuel (SNF). In Sweden, the spent nuclear fuel from all the reactor sites is moved to the Swedish central interim storage facility for spent nuclear fuel (for which the Swedish acronym is Clab). A new facility, Clink, is planned at the site where the SNF will undergo a safeguards verification prior to encapsulation for long-term storage. The fuel to be encapsulated includes both regular fuel assemblies as well as “non-regular” fuel assemblies including fuel objects called quivers, which are specially designed containers to house damaged or failed and leaking spent fuel rods in a way to isolate the rods from the environment and prevent contamination. The quiver concept was recently introduced in the Swedish nuclear market by Westinghouse Electric Sweden AB and it has led to some unique challenges from a safeguards verification standpoint which stem from their construction. Their overall stainless steel build, while providing robustness to the structure, also greatly diminishes the possibility of detecting gamma or neutron radiation using traditional safeguards measurement devices. The current investigation looks into the practicalities of safeguards verification of boiling water reactor (BWR) quiver objects in the spent fuel pool from above, and also assesses the possibility of their verification from the side using the widely used Fork detector. The Fork instrument has been routinely employed by both operators and inspectors around the world to verify spent fuel for routine safeguards inspections. In the present work, we model the BWR quiver and the Fork instrument in the Monte Carlo particle transport code, Serpent2 to estimate the radiation flux around the quiver objects. We have shown that the gamma and neutron radiation from the BWR quiver were heavily attenuated by the stainless steel lid and could not be relied on to make a safeguards verification from above. Furthermore, it was established that while gamma radiation from the quiver remains measurable on the sides of the quiver by the Fork instrument, the neutron counts were low compared to a typical BWR fuel assembly of similar fuel content albeit within the limits of detectability of the Fork.

Time stamped list mode data from gamma-ray spectroscopic measurements on 47 nuclear fuel assemblies performed at Clab, Sweden, September 2016 through March 2019

Using a high-purity Germanium gamma-ray energy spectroscopic detector system, time-stamped list-mode data sets were acquired during axial scanning of 19 boiling water reactor (BWR) and 28 pressurized water reactor (PWR) type of nuclear fuel assemblies.The data sets were collected during two measurements campaigns in September 2016 and March 2019 at the Central Interim Storage Facility for Spent Nuclear (Clab) in Sweden. A certified calibration source of 137Cs was positioned along the central line of sight between the measured fuel assembly and the detector. Data sets from measurements with only the calibration source and other background sources, i.e. without a nuclear fuel assembly present, are also included.The list-mode structure of the measured data allows for an axially-resolved as well as energy-spectral resolved intensity of nuclide-specific gamma lines emitted from the spent nuclear fuel. Data presented here can be used e.g. for validation of gamma-ray transport simulation tools or for development of methods to estimate parameters of the spent nuclear fuel based on data from gamma-ray spectroscopy.

Advancing the Fork detector for quantitative spent nuclear fuel verification

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms. The results are summarized, and sources and magnitudes of uncertainties are identified, and the impact of analysis uncertainties on the ability to confirm operator declarations is quantified.

Experimental validation of depletion calculations with VESTA 2.1.5 using JEFF-3.2

The removal of decay heat is a significant safety concern in nuclear engineering for the operation of a nuclear reactor both in normal and accidental conditions and for intermediate and long term waste storage facilities. The correct evaluation of the decay heat produced by an irradiated material requires first of all the calculation of the composition of the irradiated material by depletion codes such as VESTA 2.1, currently under development at IRSN in France. A set of PWR assembly decay heat measurements performed by the Swedish Central Interim Storage Facility (CLAB) located in Oskarshamm (Sweden) have been calculated using different nuclear data libraries: ENDF/B-VII.0, JEFF-3.1, JEFF-3.2 and JEFF-3.3T1. Using these nuclear data libraries, VESTA 2.1 calculates the assembly decay heat for almost all cases within 4% of the measured decay heat. On average, the ENDF/B-VII.0 calculated decay heat values appear to give a systematic underestimation of only 0.5%. When using the JEFF-3.1 library, this results a systematic underestimation of about 2%. By switching to the JEFF-3.2 library, this systematic underestimation is improved slighty (up to 1.5%). The changes made in the JEFF-3.3T1 beta library appear to be overcorrecting, as the systematic underestimation is transformed into a systematic overestimation of about 1.5%.

Passive gamma analysis of the boiling-water-reactor assemblies

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat.This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE

The energy release rate from the decay of radionuclides can be a critical design parameter for used nuclear fuel storage, transportation, and repository engineered systems. Validation of the SCALE nuclear analysis code system capabilities in predicting decay heat for commercial used fuel applications has been performed using decay heat measurements for fuel assemblies irradiated in pressurized and boiling water reactors. The experimental data used for validation include a large number of full-length-assembly decay heat measurements that were performed between 2003 and 2010 at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, Clab, operated by the Swedish Nuclear Fuel and Waste Management Company, SKB. The measured fuel assemblies cover the burnup range 14–51GWd/MTU and cooling times between 12 and 27 years, which are times of interest to used fuel transportation and storage applications. The validation results indicate good agreement between calculated and measured decay heat values, generally within the reported measurement uncertainty. The effects of key modeling assumptions and data used in the calculations are presented and discussed.

Comparison of the transportation risks for the spent fuel in Korea for different transportation scenarios

According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. At the start of the operation of the final repository (FR), by the year 2065, transport will then take place between the CISF and the FR. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios for a maritime transportation by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. And, we estimated and compared the transportation risks for these four transportation scenarios. Also, we estimated and compared the transportation risks resulting from accidents during the transportation of PWR and PHWR spent fuels by road trailers from the CISF and the FR. From the results of this study, we found that risks resulting from accidents during the transportation of the spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future.

SCALE analysis of CLAB decay heat measurements for LWR spent fuel assemblies

A set of decay heat measurements for spent fuel assemblies recently carried out at the Swedish central interim storage facility for spent fuel, CLAB, was analyzed with the SCALE code system. The measurements include a variety of light water reactor assemblies that cover a large burnup range – up to 51 GWd/MTU – and a cooling time domain of interest to spent fuel storage and transportation applications. The results of the analysis show a good agreement between measured and predicted decay heat, with the calculated decay heat in general within the range of the uncertainty of the measured value. The effect of various assembly data on the calculated decay heat is analyzed and discussed. Uncertainties that may arise from various approaches and assumptions in the computational model are identified and examined.

Dosimetry at the Interim Spent Fuel Storage Facility of the Czech Nuclear Power Plant Dukovany

Dosimetric characteristics of neutron and photon components of mixed fields have been determined at various places of the dry interim storage for spent nuclear fuel situated at the Nuclear power plant Dukovany. The results obtained with metrological grade instruments were compared with data provided by usual survey meters for both neutrons and photons. Neutron spectra measured at Dukovany are compared with those got at the CLAB - Central interim storage facility Oskarshamn, Sweden.

Progress towards a Deep Repository for Spent Nuclear Fuel in Sweden

The Swedish Nuclear Fuel and Waste Management Co. has in operation a safe and well integrated system for handling of all radioactive residues within Sweden. The existing central repository for low- and medium-level waste (SFR) and the central interim-storage facility for spent nuclear fuel (CLAB) can accommodate all the radioactive waste produced inside Sweden. Comprehensive research, development and demonstration activities are well under way for an encapsulation plant and a deep repository for spent fuel. These two facilities remain to be constructed to complete the waste management system. Siting of the deep repository is in progress with the aim of finding a suitable and accepted site. Implementation of the deep geological repository is a technical, scientific, social and political challenge. Smooth implementation must take into consideration both facts and emotions. Patience, flexibility and respect for the democratic process are important keywords. Research facilities, such as the underground Äspö Hard Rock Laboratory and the Encapsulation Laboratory, are important to promote scientific understanding as well as to demonstrate the disposal concept and technology.

Progress towards a Deep Repository for Spent Nuclear Fuel in Sweden.

The Swedish Nuclear Fuel and Waste Management Co. has in operation a safe and well integrated systen for handling of all radioactive residues within Sweden. The existing central repository for low- and medium-level waste (SFR) and the central interim-storage facility for spent nuclear fuel (CLAB) can accommodate all the radioactive waste produced inside Sweden. Comprehensive research, development and demonstration activities are well under way for an encapsulation plant and a deep repository for spent fuel. These two facilities remain to be constructed to complete the waste management system. Siting of the deep repository is in progress with the aim of finding a suitable and accepted site. Implementation of the deep geological repository is a technical, scientific, social and political challenge. Smooth implementation must take into consideration both facts and emotions. Patience, flexibility and respect for the democratic process are important keywords . Research facilities, such as the underground Aspo Hard Rock Laboratory and the Encapsulation Laboratory, are important to promote scientific understanding as well as to demonstrate the disposal concept and technology.