Full Core Research Articles

Abstract LB373: Pilot whole-exome sequencing data of formalin-fixed paraffin embedded endometrial tumors in the Epidemiology of Endometrial Cancer Consortium (E2C2)

Abstract Background: Endometrial cancer (EC) is the most common gynecologic cancer in the U.S.: incidence and mortality are rising. Black women face 2-fold higher mortality compared with other racial groups. To understand the role of tumor biology underlying these health inequities, we are sequencing the tumor exomes of 1,000 Black and 2,000 non-Black women in the Epidemiology of Endometrial Cancer Consortium. To inform this effort, we have completed whole-exome tumor sequencing in 35 EC patients. We present early sample performance and mutational insights of this pilot. Methods: We extracted DNA from formalin-fixed paraffin-embedded (FFPE) tumor tissues (cores, scroll) and matched normal specimens (blood, tissue) from 35 women diagnosed with EC in the Louisiana Tumor Registry, Nurses’ Health Study, and Polish Endometrial Cancer Study. Whole-exome sequencing was conducted by the Center for Inherited Disease Research at Johns Hopkins University, which called somatic alterations using matched normal samples as the germline profile (NovaSeq6000). We excluded two contaminated core samples and one paired scroll with extremely low coverage, leaving 33 women in our analytic sample. We examined the mutational landscape of EC overall and separately in 7 Black and 26 non-Black patients. For 23 women, we obtained both an FFPE scroll of the full block and FFPE tumor cores enabling a direct comparative analysis of DNA extraction yield and sequencing performance by sample source. We ran a published allele-specific copy number and clonal heterogeneity analysis tool (FACETS) in these 23 women to calculate tumor purity by tissue type. Results: Results from our pilot study revealed the most frequently mutated genes in our pilot population were PTEN (61%), ARID1A (48%), and PIK3CA (39%). We observed similar mutation frequencies for PIK3CA in both non-Black and Black women (38% vs 43%). TP53 was more frequently mutated in Black women (57% vs 16%), while ARID1A mutation was more common in non-Black women (58% vs 14%). 10 (30%) patient samples had chromosome 10 copy-neutral loss of heterozygosity. Of these, 8 (80%) had PTEN bi-allelic alterations. Of participants with both core and scroll tissue available, both sample types were successfully sequenced, although cores showed modestly better performance: 74% of core samples (vs 26% for scrolls) had a higher percentage of bases read at a target depth of 100x, and the median tumor purity for cores estimated using FACETS was higher than scrolls (53% vs 42%). Tumor purity could not be calculated for seven tissue samples due to either tumor diploidy or purity being too low: of these samples the majority were scrolls (86%). We observed greater concordance between core and scroll variant allele frequency when tumor purity was ≥30%: the benefits of cores vs scrolls matter at lower tumor purity. ConclusionsIn line with TCGA data, these pilot findings suggest potential differences in tumor mutational landscape by race and we will be interrogating these in our larger study population. FFPE core tissue generally performed better than FFPE scroll tissue in whole-exome sequencing, but our data suggest scrolls remain a viable option for sequencing when cores are unavailable. This has the potential to offer valuable insight for conducting population-based next-generation studies, which mostly rely on FFPE-derived tissues. Citation Format: Noah C. Peeri, Megan A. Clarke, Immaculata De Vivo, Mengmeng Du, Lucio Miele, Jeffin Naduparambil, Kathryn L. Penney, Venkat Seshan, Ronglai Shen, Veronica W. Setiawan, Nicolas Wentzensen, Xiao-Cheng Wu, Arnold Zea, Wenyu Zhang. Pilot whole-exome sequencing data of formalin-fixed paraffin embedded endometrial tumors in the Epidemiology of Endometrial Cancer Consortium (E2C2) [abstract]. In: Proceedings of the American Association for Cancer Research Annual Meeting 2024; Part 2 (Late-Breaking, Clinical Trial, and Invited Abstracts); 2024 Apr 5-10; San Diego, CA. Philadelphia (PA): AACR; Cancer Res 2024;84(7_Suppl):Abstract nr LB373.

Broadband low-frequency sound insulation of stiffened sandwich PFGM doubly-curved shells with positive, negative and zero Poisson's ratio cellular cores

Compared to the conventional cellular cores, the cellular cores auxetic metamaterials with negative and zero Poisson's ratios have unique mechanical deformation characteristics, which can be further applied to modeling lightweight sandwich structures. Based on that, the sound transmission characteristics of a novel stiffened sandwich porous functionally graded materials (PFGM) doubly-curved shells with full Poisson's ratio characteristic range cellular cores are investigated in this study. The adjustment of material properties of PFGM face sheets in the thickness direction is achieved through power law based on component volume fraction, and the core layer consists of cellular cores with positive, negative, and zero Poisson's ratios (PPR, NPR, and ZPR). Two common types of stiffened sandwich shell types, namely, the shell with PFGM face sheets and metal-rich cellular cores (type-A), and the ceramic-rich cellular cores (type-B), are discussed, and the Hamilton's principle is used to derive the governing equations. By applying normal velocities continuity conditions at the fluid-structure interface to consider fluid-structure coupling, which is further analytically solved using Navier's technology, and the average sound transmission loss (STL) with broadband low-frequency is described. The effectiveness of the established theoretical model is confirmed by comparing it with previously published results, experimental and simulated results obtained by impedance tube and COMSOL commercial software. A systematic comparison is conducted on the performance of stiffened sandwich PFGM doubly-curved shells with full Poisson's ratio characteristic range cellular cores, and the results show superior performance in the sound insulation for the sandwich shell with ZPR cellular cores, considerably at a broad low-frequency range of 220–2040 Hz, in comparison with the sandwich shell with NPR cellular cores and conventional PPR cellular cores types. This work's findings can serve as a benchmark for future studies and help to design and create engineering sandwich structures with improved sound insulation properties.

Availability and Geographic Access to Hospital-Based Breast Cancer Diagnostic Services in Ghana.

Breast cancer is the most frequent cancer and second most common cause of cancer-related death in Ghana. Early detection and access to diagnostic services are vital for early treatment initiation and improved survival. This study characterizes the geographic access to hospital-based breast cancer diagnostic services in Ghana as a framework for expansion. A cross-sectional hospital-based survey was completed in Ghana from November 2020 to October 2021. Early diagnostic services, as defined by the National Comprehensive Cancer Network (NCCN) Framework for Resource Stratification, was assessed at each hospital. Services were characterized as available >80% of the time in the previous year, <80%, or not available. ArcGIS was used to identify the proportion of the population within 20 and 45 km of services. Most hospitals in Ghana participated in this survey (95%; 328 of 346). Of these, 12 met full NCCN Basic criteria >80% of the time, with 43% of the population living within 45 km. Ten of the 12 met full NCCN Core criteria, and none met full NCCN Enhanced criteria. An additional 12 hospitals were identified that provide the majority of NCCN Basic services but lack select services necessary to meet this criterion. Expansion of services in these hospitals could result in an additional 20% of the population having access to NCCN Basic-level early diagnostic services within 45 km. Hospital-based services for breast cancer early diagnosis in Ghana are available but sparse. Many hospitals offer fragmented aspects of care, but only a limited number of hospitals offer the full NCCN Basic or Core level of care. Understanding current availability and geographical distribution of services provides a framework for potential targeted expansion of services.

A proposed refueling scheme for the modular pebble bed reactor

In this paper, a suggested design of the fuel pebble and loading pattern in the MIT Modular Pebble Bed HTGR (MPBR) using thorium (232Th) is presented. Using the MCNP5 code model for the MPBR reactor, different compositions of Thorium Low Enriched Uranium (ThLEU) with different Heavy Metal (HM) loading and 235U enrichment were investigated to achieve the maximum thorium utilization with relatively minimum reactivity insertion in case of water ingress accident. Using pure Low Enriched Uranium (LEU) pebbles and ThLEU pebbles with the same HM loading and 235U enrichment, different pebbles loading patterns were studied. One loading pattern was selected to achieve the criticality with enough excess reactivity compensating for the poisoning effect of xenon build-up at the startup of the reactor operation and the negative effect of fuel and reflector temperature coefficients of reactivity. This loading pattern aims also to reduce the radial power peaking in the core and to minimize the axial Power Peaking Factor (PPF) in the central hot channel, with a minimum number of pure LEU fuel batches. A calculation procedure was developed that combines fuel pebble depletion and full core calculations to investigate the impact of a proposed online refueling scheme for the MPBR reactor and to simulate the multicycle refueling mechanism. Using SERPENT 1.1.7, the nuclide concentrations were calculated as a function of pebble burnup for both LEU and ThLEU pebbles. A multicycle online refueling scheme was proposed and the average number of passes required for LEU and ThLEU pebbles to reach their maximum burnup was estimated.

Neutronic Assessment of High-Temperature Gas-Cooled Thorium Burner using Monte Carlo Calculation Method with Full Core Model

In this study, the effective reactivity and burnup analyses have been performed for heterogeneous three-dimensional high-temperature gas-cooled thorium reactor (HTGR) which has 60 MWth full core geometry by using continuous-energy multi-purpose three-dimensional Monte Carlo particle transport Serpent code with ENDF/ B-VII data libraries. Nuclear fuel have been selected as 50 % ThO2+50% RG-PuO2. Firstly, effective reactivity for three different qualities of graphite for operation period have been determined. The effective reactivity showed better performance with increasing densities of graphite. Secondly, it has been also examined to ZrC and SiC cladding materials effect on the effective reactivity. It is observed that SiC has a positive effect on reactivity compared to ZrC. As a results, the full core life low-power thorium-burner HTGR have been calculated as up to ~4500 days depending on the graphite material whereas, the corresponding burn−ups came out to be ~ 189 GWd/ton, for end of life.

The deterministic neutron transport calculation for the HTR-PM by the three-dimensional method of characteristics code ARCHER

Due to the strong geometric adaptability, the Three-Dimensional (3-D) Method Of Characteristics (MOC) is a promising candidate for 3-D whole-core high-fidelity neutron transport calculations. However, the 3-D MOC can hardly be applied to large-scale full core problems on account of the huge computational costs in memory and time. In the past, the 3-D MOC was mainly used in traditional Light Water Reactor (LWR) calculations, but rarely used for more complex core calculations. For pebble-bed High Temperature gas-cooled Reactors (HTRs), most 3-D MOC codes are out of ability to construct complex pebble beds, let alone efficient acceleration methods. The 3-D MOC code ARCHER can efficiently simulate the large-scale pebble-bed HTRs through using the Linear Source Approximation (LSA), the Coarse Mesh Finite Difference (CMFD) and the hybrid MPI-OpenMP parallel. In this work, the two-level CMFD acceleration and efficient preconditioners on Krylov subspace linear solvers are implemented in the ARCHER code in order to achieve better performance. In addition, the practical HTR-PM criticality problem is calculated by ARCHER, which is the world’s first high-fidelity neutron transport solution of large-scale pebble-bed HTRs by deterministic numerical method. It also shows that the 3-D MOC has great application potential in other complex reactors.

Transient full core analysis of PWR with multi-scale and multi-physics approach

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

Evaluation of the spatial self-shielding impact for TRISO-based nuclear fuel depletion

Reactor physics analyses of nuclear cores with nuclear fuel concepts containing tristructural isotropic (TRISO) particles, such as pebbles or compact fuel elements, rely on various degrees of simplification to keep these highly heterogeneous problems computationally tractable. One such limitation regards the level of spatial discretization employed during burnup calculations, where traditionally only a limited number of spatial zones are modeled at the full core level and assume that the spectrum is constant within the fuel elements and TRISO particles in this depletion zone. This type of assumption neglects the impact of spatial self-shielding effect within the kernels (microscale level) as well as within the compact or pebbles (mesoscale level). The Monte Carlo code Serpent 2 contains many relevant features for efficiently modeling this type of geometry, including a collision-based domain decomposition intended for very large burnup calculations, which we leveraged for this work to quantify the impact of capturing neutron flux variations occurring at the micro- and mesoscale level on a series of high-temperature gas-cooled reactor fuel element depletion problems. While spatial self-shielding is observed at both scales, with differences from a volume-averaged burnup of ±7% within the kernels and ±2% between TRISO particles within the fuel element, the conjugated effect on nuclide inventories and multiplication factor are negligible, hence confirming that assuming a single average spectrum value may be sufficient for most applications.

Cross section generation for a conceptual horizontal, compact high temperature gas reactor

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

Virtual lattice method for efficient Monte Carlo transport simulation of dispersion nuclear fuels

The randomly dispersed Tristructural-isotropic (TRISO) particle fuel is an attractive nuclear fuel type widely used in advanced reactor designs. However, the massive randomly-distributed fuel elements and the prohibitive computational costs pose challenges for high-fidelity modeling and simulation of dispersion fuels. To address this, we present a virtual lattice optimization method for accelerating the Monte Carlo particle transport simulation of nuclear systems with dispersion fuels. The new method involves using a regular overlaid mesh on the dispersed fuel region to improve geometry processing in Monte Carlo simulations. We describe the optimization methodology in our paper and construct a theoretical performance model to determine the optimal lattice pitch size that minimizes calculation cost. We then develop and implement the algorithm of the virtual method in the open-source Monte Carlo code OpenMC. To demonstrate the efficacy of the new algorithm, high-fidelity HTR-PM full core models based on the Shidao-Bay nuclear power plant are constructed, including an explicit representation of up to 420,000 fuel pebbles. Criticality and depletion simulations are performed on large-scale models. Results show that the k-effective derived by the optimized code agrees well with both the experiment and the original code. Further performance comparison shows that the virtual lattice method exhibits lower time consumption, higher parallel efficiency, and lower storage requirement in comparison to the conventional physical lattice method. The study establishes the effectiveness of the proposed methodology in facilitating efficient and accurate high-fidelity simulations of large-scale dispersion fuel models.

The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0

Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core “pin-by-pin” is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis.

Thermal-hydraulic modelling and stade staty analysis of a LS-VHTR reactor

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) is a liquid salt cooled and TRISO fueled reactor, which presents very good features in terms of power production and safety aspects. The advantage of using liquid salt as coolant is mainly related with its high efficiency of heat transfer and operation at low pressures. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core, in steady state conditions, starting from the single channel level and concluding with the full core model. The LS-VHTR is a new reactor concept that introduces some different features in relation to other nuclear reactors. In this way, a thermal modeling of this reactor has been developed using the RELAP5-3D code using a point kinetic model. A modeling with three channels has been presented. The core coolant temperature increase, the core coolant mass flow and the pressure along the core were simulated and presented in good agreement with the expected behavior

Transport equivalent full core neutronics calculations via finite volume method and optimized diffusion parameters

In this study, three-dimensional neutronics analysis of the hot zero power (HZP) situation of the BEAVRS full core benchmark problem is performed by employing a finite volume method (FVM), Krylov subspace methods, and transport equivalent diffusion model (TEDM). TEDM is used for generating transport-corrected diffusion parameters and we use a genetic algorithm for the correction. The study shows that a transport equivalent and accurate neutronics analysis of a highly detailed commercial light water reactor (LWR) core is feasible using only the two-group diffusion theory. The core model in the benchmark is represented by 165 different diffusion parameter sets since the model is highly heterogeneous. The reaction rates, flux, and current distributions obtained from Serpent code, the fuel assembly-based FVM diffusion theory solutions, and a genetic algorithm are used to generate the transport equivalent diffusion parameters. The task of the genetic algorithm used is to minimize the difference between Serpent and diffusion theory solutions by estimating the two group diffusion coefficients and the scattering matrix. The transport equivalent diffusion parameters are then used for the full core neutronics analysis by FVM and Krylov subspace methods. The results of the analysis are compared with the measured data provided by the BEAVRS benchmark and the results of Serpent code simulations and found to be in complete agreement with the Serpent code results and satisfactorily consistent with the measured data provided by the BEAVRS benchmark.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Ultrasound Contrast Agent Priming of Biopsy and Introducer Needles by Using a Small Syringe to Improve Needle Visibility in a Phantom Model

PurposeBiopsy under the guidance of contrast-enhanced ultrasound is sometimes useful. Needle visualization in contrast-specific imaging-mode is often poor; however, it may be improved by priming the needles with an ultrasound contrast agent. This study aimed to evaluate needle priming methods using the ultrasound contrast agent sulfur hexafluoride and a 1 mL syringe.Material and MethodsTwo kinds of biopsy needles, side-notch and full core, and one kind of introducer needle were primed using non-primed needles as controls (n = 180). Recordings of punctures were performed in a water bath phantom to which the ultrasound contrast agent had also been added. Contrast-specific imaging-mode needle visibility was evaluated for the entire needles and the needle tips, respectively, quantitatively by calculating the contrast-to-noise ratio and qualitatively via grading by three radiologists.ResultsThe contrast-to-noise ratio following the ultrasound contrast agent priming was superior compared to the controls for the entire needles of all three types (p < 0.001) and for the needle tips of the core biopsy needles and introducer needles (p < 0.001). However, the ratio was equal to the controls for the needle tips of the side-notch biopsy needles (p = 0.19). Needle visibility following the ultrasound contrast agent priming was qualitatively superior compared to the controls for both the entire needles and the needle tips, and the difference was considered clinically relevant by the assessors (p < 0.001).ConclusionThe ultrasound contrast agent needle priming methods described increased the contrast-specific imaging-mode needle visibility in a phantom model. Nonetheless, the results also need to be confirmed in vivo.

Digital rock modeling to quantify scale dependence of petrophysical measurements in burrowed reservoir rocks: An example using Thalassinoides

Recent evaluations of burrow connectivity have shown that burrow morphology, abundance, and size impact the sampling scheme required for faithfully representing permeability in the laboratory. Quantifying these controls to design appropriate sampling is essential. This study illustrates a new approach using multipoint statistics modeling and applies it to Thalassinoides, to simulate a range of burrow abundances and sizes, and then interrogates the outcome to develop a statistical model that allows sampling strategies to be designed. Our workflow started with digital models that were known to have open or permeable Thalassinoides that connected one side of the model volume to another, (i.e., high permeability). We cut the model volumes into 540 digital samples with a range of burrow abundance, burrow size and sample cross section. The Thalassinoides showed no cross-sample connectivity in 173 (32%) samples; the remaining 367 (68%) showed cross-sample Thalassinoides connectivity. The data on burrow abundance, burrow size, and sample dimensions (as independent variables) and the result of samples representing Thalassinoides connectivity or not (as a dependent variable), was analyzed using logistic regression modeling. An important outcome from the logistic regression model is an equation that provides a probability of Thalassinoides connectivity across samples, given burrow abundance, burrow size, and sample dimension inputs. This equation has practical applications in determining the minimum sample cross-sections and sample lengths needed for samples to adequately represent the permeability. As examples, for samples with cross sections similar to core plugs and full diameter core, given various burrow percentages and sizes, only 37% yield burrow connectivity across any and all sample lengths. For the rest, our data on the largest connected burrow length allows for calculation of the probabilities of samples of typical lengths being representative of the burrow connectivity. Thus, researchers can assess the likelihood of their samples representing Thalassinoides connectivity given a range of burrow abundance, burrow size, sample cross section and sample length. The spreadsheet to make these calculations and lookup tables are provided as digital accessory files. The example using Thalassinoides provides proof-of-concept of a methodology that should be broadly applicable to any given burrow morphology. An algorithm and lookup table that can be modeled for each given burrow morphology with an appropriate training image.

Improvement on Fourier fundamental mode coefficient for Monte Carlo criticality calculations

Recently, the Fourier fundamental mode coefficient (FFMC) was formulated to advance the existing Monte Carlo fission source stationarity diagnostics. The calculation of FFMC requires a user-defined exterior geometric boundary of the fission source distribution. However, the effect of using different boundary sizes on the reliability of the proposed FFMC indicator is not understood and has never been discussed. This work used different boundaries to investigate their effect on the FFMC results based on a simple homogeneous reactor model. The results show that the choice of FFMC boundaries is also affected by the reactor model's initial source distribution and geometrical configuration. The shortcoming of the abovementioned indicator has been improved by dissociating the indicator into the corresponding Cartesian axes. The new set of indicators is further verified by the checkerboard fuel storage array and the BEAVRS full core cases.

Improving source convergence diagnostics sensitivity using the complex components of Fourier fundamental mode

Previous works have demonstrated that the Fourier fundamental mode coefficient (FFMC) is a potential fission source convergence indicator for Monte Carlo k-eigenvalue calculations. Unlike the conventional Shannon entropy indicator, it requires no spatial meshing, and its computation is computationally inexpensive. However, we found that FFMC is less sensitive to the convergence involving near symmetrical fission distributions. To rectify the problem, we focus on the analysis of the FFMC’s components (the real and imaginary parts) on a simple geometry problem and a full reactor core problem. The usage of the FFMC’s components is more conservative compared to the original FFMC method as supported by the improved results of the fission source convergence detection.

Coupled module of radioactive waste analysis for CANDU reactor based on the Monte Carlo method

When decommissioning nuclear power plants, it is necessary that various types of radioactive wastes are to be estimated and be analyzed accurately. The first decommissioning process of the commercial light water reactor (LWR) is expected to start soon in South Korea, however, lots of processes are still under development for CANada Deuterium Uranium (CANDU) reactor. Unlike the LWR type reactor, the CANDU has complicated and different shapes of components in the reactor, which makes it difficult to carry out activation analysis considering region-wise different irradiation histories. In this study, a coupled module has been developed with Monte Carlo full core analysis and the region-wise depletion analysis by equipped automatic input generation. Furthermore, the levels of radioactive wastes are determined for transportation including estimation of cask contents based on the coupled module results.

Definition of the neutronics benchmark of the NuScale-like core

This paper defines a 3D full core neutronics benchmark which is based on the NuScale small modular reactor (SMR) concept. The paper provides a detailed description of the NuScale-like core, a list of expected outputs, and a reference solution to the benchmark exercises obtained with the Monte Carlo code Serpent.The benchmark was developed in the framework of the Euratom McSAFER project and can be used for verification of computational chains dedicated to 3D full-core neutronics simulations of water cooled SMRs.The paper is supplemented with a digital data set to ease the modeling process.