Fissile Mass Research Articles

A computational framework to support probabilistic criticality modelling for the geological disposal of radioactive waste

Nuclear Waste Services is tasked with disposal of the UK’s higher-activity radioactive waste in a Geological Disposal Facility. The disposal of fissile nuclides requires a demonstration that there is no significant concern from criticality, i.e. a fission chain reaction. While waste packages will initially be emplaced in a subcritical configuration, over the long timescales following closure there is potential for waste packages to degrade and for nuclides to be dispersed in the subsurface by groundwater, leading to the potential for a critical system forming. To facilitate modelling, a codebase has been developed which interfaces a probabilistic simulation tool (GoldSim) with a neutron transport code (MONK/MCNP). This allows large ensemble simulations to be run iteratively to determine limiting fissile masses which satisfy a criticality safety criterion. This paper documents the main algorithms and methodologies implemented within this framework, and provides background and example results illustrating the application to post-closure criticality modelling.

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Neutronic and Proliferation Resistance Analysis of Small Modular Pressurized Water Reactor with Various Fuel Types using SRAC 2006

One strategy for emission reduction in Indonesia is the dedieselization program, replacing diesel power plants with renewables. Pressurized Water Reactor (PWR) type Small Modular Reactor (SMR) is also a consideration. This study aims to compare fuel performance in SMRs. Four fuel types with varying fissile and fertile nuclides but equal fissile mass percentages were assessed: U235- U238, U235-Th232, U233-U238, and U233-Th232, using PIJ SRAC 2006 and JENDL 4.0. Parameters include neutron multiplication factor, conversion ratio, and isotope production. Results indicated that U233-U238 has a highest initial neutron multiplication factor, 1.55. Substituting uranium-238 with thorium-232 boosts neutron flux and conversion, reducing nuclear proliferation risk. While initial neutron multiplication may not be highest, thorium-232 can produce uranium-233 with high initial neutron multiplication factor. This research aids fuel selection for Indonesia SMR dedieselization program.

Modelling of spent fuel dissolution process for assessment of criticality safety margin

Criticality safety margin assessment is a challenging task for a large dissolution tank under uncertain process conditions. The Monte-Carlo based neutron transport code (MCkeff) was used to model geometry, material and process conditions of dissolution tank containing fissile material. This manuscript presents a sampling-based probabilistic technique for estimating single-sided tolerance limit on criticality safety parameter due to random variation of process parameters. Based on their effect on fissile mass, fissile concentration, neutron moderation and neutron reflection, potential uncertain parameters impacting the criticality safety margin were identified. By withdrawing random samples from uncertain domain of process parameter, a simulation test matrix was formed to study the variability of the neutron multiplication factor. The Wilk's formula, which corresponds to the 95th percentile and 95% confidence level, was used to estimate required sample size to minimise statistical error (i.e., Type-I error associated to the rejection of the genuine sample and Type-II error connected to the acceptance of the false sample). In order to cut down the computation costs associated with requirement of large computer runs, a meta-model for the criticality safety problem was also constructed using a linear regression model. Additionally, associated model uncertainty was quantified. The variation in neutron multiplication factor caused by changes in process variables like pellet diameter, linear pellet mass, temperature, etc. was also investigated as an input parameter uncertainty. It is found that model uncertainty plays significant role while assessing the criticality safety margin using meta-model.

A micro-sized high-temperature thorium reactor with Duplex TRISO fuel with enhanced thorium utilization compared to the SBU configuration

This study offers a neutronisc core calculation to assess the potential of the compact microheterogeneous thorium-uranium TRistructural ISOtropic (TRISO) fuel design for the micro-modular high temperature reactor. The Duplex-fueled core and the Seed-and-Blanket-Unit (SBU)-fueled core were modelled and compared using the MCNPX 2.7 Monte Carlo code. The Th:U ratio and initial fissile mass of both cores were set to be the same. The dispersion of thorium and uranium in the Duplex-fueled core is more uniform compared to that in the SBU-fueled core. The study demonstrates that Duplex TRISO fuel offers both improved core power distribution uniformity and superior neutronic performance compared to SBU fuel in a core. The utilisation of a Duplex-fueled core leads to a 5 % higher core conversion ratio and resonance capture rate, which ultimately results in an extra long-term production of 233U. Additionally, its FTC value is lower. The required material and thickness of the neutron reflector, are also discussed, though only as a preliminary analysis. Nevertheless, the findings clearly demonstrate that the Duplex TRISO fuel design concept offers valuable flexibility and optimization in the distribution of fertile and fissile material mass. These characteristics are of utmost importance for the advancement of thorium reactors, as they significantly enhance neutronic performance.

Fissile mass estimation using β-delayed γ-rays and neutrons from fast and epithermal induced fissions

A potential method for mass estimation of shielded special nuclear materials (SNM) is proposed. The method uses fast and epithermal neutrons, generated by an external pulsed neutron source, to induce fission in SNM. The subsequent β-delayed γ-rays and neutrons, emitted by the fission products, are measured simultaneously by two 3He slab detectors and a coaxial high purity germanium (HPGe) detector, located around the tested material. Measuring simultaneously both, β-delayed γ-rays and neutrons measurement, offers improved sensitivity for fissile materials detecting.Laboratory examination of the method was performed in the Pulsed Neutron Interrogation Test Assembly (PUNITA), using CBNM (Central Bureau for Nuclear Measurements) samples of Uranium.

Transient nuclear criticality excursion analysis of highly dispersed particulate three-phase fluidised systems

The aim of this study was to perform sensitivity analysis, investigating how different fluidisation and sedimentation characteristics of three-phase wetted UO2 powder beds, may affect a transient nuclear criticality excursion initiated through the addition of water into a fissile powder bed. This type of postulated nuclear criticality accident scenario may occur in nuclear fuel fabrication facilities when a fire is fought through the use of water, supplied via an automatic or manual fire-suppression system. A similar scenario may also develop as a result of water leaks or flooding of the process area housing UO2 powder. The article introduces a model for gas-bubble induced fluidisation of a UO2 powder bed and examines how this phenomenon may affect the neutron kinetic response of a three-phase fluidised fissile powder system. Empirical analysis has shown that fissile suspensions form agglomerated structures when suspended in water, at agglomerate sizes that range from 18 μm to 40 μm. Simulation results indicate that both the critical gas velocity and rate of fluidisation may significantly affect transient nuclear criticality excursion dynamics. The re-distribution of fissile mass into a highly dispersed suspension generally reduces the reactivity of the system, however, depending on the H/U ratio, a positive reactivity may be added to the system. Low Péclet numbers in the suspension suggest that gas-bubble induced motion of the suspension causes a highly dispersive flow field. An oscillatory power response is predicted for low critical gas velocities where the reactivity of the system is predominantly governed by the re-distribution of fissile mass within the system. The frequency of these oscillations is greater for a higher hindered settling rate of powder particles. At a higher critical gas velocity, the transient nuclear criticality excursion is governed by the voidage reactivity feedback, making the response quite independent of fluidisation. In all cases, large volumes of UO2 powder may leave the domain due to overflowing of the suspension. Transient nuclear criticality excursions in UO2 powder beds with a low critical gas velocity are terminated once the bed becomes fully saturated.

Improving FNMC for the matrix effect of spherical shell plutonium samples

Abstract The fissile mass deduced from fast neutron multiplicity counting (FNMC) measurement is underestimated if the matrix self-absorption effect of the radioactive source is not taken into account. Based on the analysis of FNMC equations, a set of FNMC system was built to simulate and study the mass attribute of the hollow sphere (spherical shell) plutonium under different shapes and different masses conditions. Geant4 simulation shows that an appropriate parameter correction successfully removes the bias because of the matrix effect. Consequently, the self-multiplication factor, α coefficient and scattering crosstalk of the simulation result were corrected after analyzing the detection efficiency and multiplicity counting rate, and the corresponding polynomial fitting equation was obtained. The corrected mass deviation of samples was less than ±1% in this interval. The results show that the combination of the FNMC and parameter correction can accurately measure the sample mass attribute, which provides a new method for solving similar problems.

Boron-Coated Straws Imaging Panel Capability for Neutron Emission Computed Tomography for Source Localization Inside Radioactive Drums

The evaluation of fissile mass inside radioactive waste drums by neutron measurements is essential for radioactive waste management, nuclear safety, and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by the matrix effects. Therefore, an imaging panel proposed by Proportional Technologies Inc., and composed of seven boron-coated straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated with plutonium or uranium position inside radioactive waste drums. In a previous work, a numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a ²⁵²Cf source. In the first section, the feasibility of neutron emission tomography by a setup composed of five extended imaging BCS panels is demonstrated by numerical Monte Carlo simulation. The second section details the experimental validation of the neutron emission tomography. Measurements are carried out with AmBe and ²⁵²Cf sources located inside an empty 118 L drum by rotating the BCS imaging panel around it. Afterward, the Richardson–Lucy deconvolution algorithm is applied to provide 2-D neutron source images for each angle. Finally, the 3-D images are reconstructed using the reconstruction toolkit (RTK) backprojection projection routine. The results demonstrate the capability of the BCS imaging to provide the 3-D location, i.e., axial and radial positions of one and two neutron sources. Furthermore, the first tests with this passive neutron measurement system show a satisfactory 3-D reconstruction for ²⁵²Cf and AmBe sources separated by 20 cm. Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated with plutonium or uranium localization in neutron measurements. Work is undergoing to assess the capability of this system for 118 L drums filled with organic and metallic matrices. Additionally, further prospects concern the performance of other deconvolution and reconstruction algorithms.

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.

Transport Calculation of the Multiplicity Moments for Cylinders

In a previous paper by Pázsit and Pál [“Multiplicity Theory Beyond the Point Model,” Ann. Nucl. Energy, Vol. 154 (2021)], a general transport theory calculation of the factorial moments of the number of neutrons emitted spontaneously from a sample was elaborated. In contrast to the original derivations by Hage and Cifarelli [“On the Factorial Moments of the Neutron Multiplicity Distribution of Fission Cascades,” Nucl. Instrum. Meth. Phys. Res. A, Vol. 236 (1985)] and Böhnel [“The Effect of Multiplication on the Quantitative Determination of Spontaneously Fissioning Isotopes by Neutron Correlation Analysis,” Nucl. Sci. Eng., Vol. 90 (1985)], also referred to as the point model, in the transport model the spatial and angular dependence of the internal fission chain is taken into account with a one-speed transport theory treatment. Quantitative results were given for a spherical item, and the bias of the point model regarding the estimation of the fission rate as compared to the more exact space-dependent model was estimated as a function of the size of the sphere and the factor. In the present paper the formalism and the quantitative work are extended to the treatment of items with cylindrical shapes, which are more relevant in many practical applications. Results are presented for both square cylinders () and for tall () and flat () cylinders. This way the differences between the cylinder and the sphere on one hand and those between the various cylinder shapes on the other hand can be estimated. The results show that the bias depends on the geometry of the cylinder quite moderately, but similarly to the case of the sphere, the bias of the point model is quite significant for larger item sizes and values, and it is nonconservative (underestimates the fissile mass) as well.

Active single-shot interrogation of natural uranium in industrial radioactive waste using plasma focus device

An Active Interrogation by Delayed Neutron Counting (AIDNEC) system for the assay of natural uranium in industrial waste in form of civil debris has been developed using the plasma focus (PF) device as a single-shot neutron source. The methodology of the technique involves irradiation of fissile mass by neutrons from a PF device and counting the delayed neutron emitted as a result of induced fission. In-house developed medium energy (17 kJ) PF device and 3He detector banks have been used for neutron irradiation of waste samples containing fissile masses and counting of the delayed neutrons respectively. The detector geometry and the moderators for the source and delayed neutrons have been optimized by Monte Carlo simulations. In this work, the results of these simulations and experimental observations are presented. The industrial radioactive waste samples used in this work were in the form of sand, bricks, chunks of concrete, etc. having a sample mass of few tens of kg. With D2 gas operated PF device (average yield (2.1 ± 0.8) ×108 neutrons/shot) the detection limit is determined to be 170 g of natural uranium, i.e., 1.2 g of 235U. The feasibility of the technique using the same PF device operated with D–T gas (14.06 MeV neutrons) has been assessed by Monte Carlo simulations. As D–T fusion provides two orders more neutron yield as compared to D2 fusion, the detection limit of the present system with D–T gas is estimated to be 3.3 g of natural uranium in ∼40 kg waste. Considering this low detection limit, this technique can be used for the assay of nuclear material in all types of industrial radioactive waste produced in the fuel fabrication/handling or similar facilities.

Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility

The Micro-Pocket Fission Detector (MPFD) is a small-form-factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility located at Idaho National Laboratory. The neutron and gamma-ray flux profiles and magnitudes were simulated using MCNP in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations inside the B-8 position 55 for three separate orientations of the nearby hafnium outer shim control cylinders and at a power level of 700 W(thermal). The resulting neutron and gamma-ray flux values were used to determine the MPFD response for various fissile masses and detector gas pressures. The optimal gas-operating pressure was determined to be between 30 and 60 psig. The required fissile layer mass was determined to be between 0.5 to 1.0 µg of 235U. Additionally, the gamma ray to fission fragment interaction rate was determined to be 1.43 × 103 with average energy deposition for gamma rays and fission fragments in 30 psig argon gas to be 1 keV and 3.5 MeV, respectively.

Multiplicity theory beyond the point model

Passive methods of nuclear safeguards determine the important parameters of an unknown sample from the statistics of the detection of the neutrons emitted from the item. These latter are due to spontaneous fissions and (α,n) reactions, enhanced by internal multiplication before leaking out. Based on the original work of Böhnel, the methodology of traditional multiplicity counting is based on the first three factorial moments of the number of neutrons, emitted from the sample due to one source event. These “Böhnel moments” were derived in the so-called “point model”, in which no space-dependence is accounted for, rather a uniform first collision probability is assumed for each neutron, irrespective of the position of its birth and its velocity direction, and, more important, it is assumed to be the same for all generations in the fission chain as for the source neutrons. The purpose of the present work is to derive the same factorial moments in a one-speed space-dependent model, in which the position and direction of the neutrons is accounted for, but (similarly to the original Böhnel model), no energy dependence is assumed. The integral equations for the moments are solved numerically with a collision number expansion. It is shown that compared to the space-dependent calculations, the unfolding method using the point model underestimates the fissile mass and the underestimation increases with increasing both of fissile mass and the value of α.

Localization of nuclear materials in large concrete radioactive waste packages using photofission delayed gamma rays

The characterization of radioactive waste packages is mandatory for their transport, interim storage and final disposal. In this framework, the Nuclear Measurement Laboratory of CEA DES IRESNE Institute, at Cadarache, France, uses a high-energy electron linear accelerator (LINAC) to produce an interrogating bremsstrahlung beam with endpoint energies ranging from 9 to 21 MeV to perform X-ray imaging and high-energy photon interrogation on large concrete packages. In particular, highenergy photon beam induces photofission reactions in both fissile (235U, 239Pu, 241Pu) and fertile (238U, 240Pu, 232Th, etc.) actinides possibly present in the radioactive waste. In order to assess their mass, we use delayed gamma rays emitted by their photofission products, which are measured with a 50 % relative efficiency High-Purity Germanium (HPGe) detector. Actinide differentiation, which is important for the fissile mass estimation, is based on the ratios of gamma rays emitted by different photofission products and requires appropriate corrections for the gamma attenuation in concrete. To this aim, we report here a localization method of point-like nuclear materials in the concrete matrix, based on the differential attenuation of several gamma rays emitted by a same photofission product. We use here the 1435.9 and 2639.6 keV lines of 138Cs, with both experimental data and MCNP numerical simulations to determine the (r,θ) coordinates of nuclear materials. Then, the depth inside the concrete matrix, which is determined with a precision of a few percent, mainly depending on counting statistics on 1435.9 and 2639.6 keV net peak areas, is used to correct for the different gamma ratios used in the actinide identification method. Experimental tests with uranium samples have been performed to validate the localization method.

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

Investigation of in-beam prompt and delayed neutron counting techniques for detection and characterization of special nuclear material

A neutron counting (NC) technique for in-beam active neutron interrogation has been developed and deployed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL). A distinguishing feature of the method is that the sample to be interrogated remains stationary during the irradiation and counting periods for the detection of prompt and delayed neutrons. The NC system permits measurements of prompt and delayed neutrons using a mono-energetic or broad-spectrum (white) thermal neutron beam. An MCNPX computational study of an NC system employing nine 3He tubes showed that a delayed neutron detection efficiency of 22% could be achieved. The presence of a sub-milligram mass of 235U could be revealed in less than ten minutes in the in-beam delayed neutron experiment incorporating a white thermal neutron beam. The technique readily differentiated between 235U and 233U isotopes by analysis of delayed neutron count rates. The lower flux of the in-beam DN experiment does not permit the trace analysis that is possible with in-core irradiation, but does permit non-destructive analysis of large samples and could prove invaluable for the initial survey of materials of unknown content and origin. The study also demonstrated that in-beam prompt neutron analysis could be the ultimate solution when the neutron source is weak, the sample is shielded, the fissile mass is small, or interrogation time is limited. Detection time was reduced to a few seconds in the examination of prompt neutrons either in single or coincidence method. This paper presents the application of an in-beam NC system using a mono-energetic and white thermal neutron beam at the NRU reactor and assesses the performance of a newly constructed portable NC system with a large neutron detection array.

Boron-Coated Straws Imaging Panel Capability for Passive and Active Neutron Measurements of Radioactive Waste Drums

The evaluation of fissile mass inside radioactive waste drums is essential for radioactive waste management, nuclear safety, and criticality issues. However, passive and active neutron measurements can be strongly impacted by the uncertainty on the neutron source position within the drum and by matrix attenuation effects. Therefore, an imaging panel proposed by Proportional Technologies Inc., composed of seven Boron-coated straw (BCS) detectors has been tested to localize neutron interactions, in view to reduce uncertainties associated with plutonium or uranium position inside radioactive waste drums. A numerical model of the imaging panel has been developed and validated from a comparison with experimental profiles obtained with a 252Cf source. A passive measurement system equipped with 12 such imaging panels has been designed by numerical simulation, in view to provide information on neutron source location in a 118-L radioactive waste drum filled with organic, metallic, or mixed organic–metallic matrices. Additionally, an experimental setup dedicated to active measurements with a D-T neutron generator has been implemented to test the imaging panel. Prompt fission neutron signals have been recorded, which is induced by thermal interrogating neutrons in fissile material samples. This article presents 2-D images indicating the position of fissile materials. Consequently, BCS imaging panels open interesting prospects to reduce the uncertainty associated with plutonium or uranium localization both in passive and active neutron measurements.

Correlating the fissile mass of standard uranium samples with delayed gamma rays from fission products

The Japan Atomic Energy Agency and the European Commission Joint Research Centre are collaborating to develop delayed gamma-ray spectroscopy (DGS) for nuclear materials for safeguards verification in reprocessing plants. In this paper, we describe DGS interrogation using the Pulsed Neutron Interrogation Test Assembly with standard samples of different 235U enrichments. By analyzing gamma-ray spectra, we reveal a linear correlation between the sample mass and both the total counts above 3.3 MeV and the peak counts of specific high-energy gamma-ray. We were able to observe, qualify and quantify specific gamma rays peak down to the depleted uranium (0.5 g 235U) mass sample. Based on this, we demonstrate that our technique is able to estimate the total fissile mass with a statistical uncertainty <2% when taking into account self-shielding and gamma self-absorption corrections. Using integrated counts above 3.3 MeV we were able to reduce the mass-dependent bias for the higher enrichments (∼3 to 4%) to <4%.

A Direct Nondestructive Assay for the Pu of U/TRU Ingot in Pyroprocessing Using 244Cm Neutron Albedo Reactivity Technique and Its Error

The transuranic (TRU) ingot is considered to be the most prominent target material of pyroprocessing in terms of safeguards since it contains almost all of the Pu of the feed spent fuel. Due to the high density, excessively high neutron emission rates, and high neutron multiplication of the U/TRU ingot, it is impractical to apply gamma-ray spectroscopy or neutron coincidence counting techniques to the quantification of the Pu content of the U/TRU ingot. Since the passive neutron albedo reactivity (PNAR) technique is known to be sensitive to the total fissile mass of target material and the uncertainty of its singles Cd ratio is independent of the accidental coincidence coming from the high neutron emission rate, the capability of the PNAR technique for the quantification of the Pu content of the U/TRU ingot has been investigated using the MCNPX code with a spent fuel library with 81 different cases of various kinds of initial enrichment, burnup, and cooling time. The MCNPX simulation results for the Cd ratio versus Pu content of the U/TRU ingot show the maximum error in the Pu mass between the linear fit and the real Pu content in the U/TRU ingot is 2.14% for 4.5 wt% initial enrichment cases. The results of this study show that the PNAR technique can be one possible method for the direct nondestructive assay for the Pu of the U/TRU ingot.