Pub Date : 2024-10-08DOI: 10.1016/j.anucene.2024.110955
Alexander Aures, Thomas Eisenstecken, Ekaterina Elts, Robert Kilger
The XSUSA method is a well-established stochastic sampling method for propagating nuclear data uncertainties through multigroup neutron transport calculations. To benefit from the advantages of Monte Carlo transport codes, namely modeling complex geometries and using continuous-energy nuclear data, an extension to XSUSA is proposed which allows perturbing continuous-energy nuclear data using multigroup nuclear data covariances. To verify the extension, sensitivity profiles of nuclear reactions are calculated via direct perturbation for the benchmark problems Jezebel, Godiva, LEU-SOL-THERM-002. The sensitivity profiles agree well with those obtained from TSUNAMI and Serpent. Secondly, the extension to XSUSA is applied to produce randomly sampled continuous-energy data libraries using the covariance libraries of SCALE 6.2. With these data libraries, samples of Serpent calculations are performed for Jezebel, Godiva, LEU-SOL-THERM-002, and the TMI-1 pin cell of the OECD/NEA LWR-UAM benchmark. For each problem, the multiplication factor uncertainty agrees well with the one from TSUNAMI.
{"title":"Nuclear data uncertainty propagation in continuous-energy Monte Carlo calculations","authors":"Alexander Aures, Thomas Eisenstecken, Ekaterina Elts, Robert Kilger","doi":"10.1016/j.anucene.2024.110955","DOIUrl":"10.1016/j.anucene.2024.110955","url":null,"abstract":"<div><div>The XSUSA method is a well-established stochastic sampling method for propagating nuclear data uncertainties through multigroup neutron transport calculations. To benefit from the advantages of Monte Carlo transport codes, namely modeling complex geometries and using continuous-energy nuclear data, an extension to XSUSA is proposed which allows perturbing continuous-energy nuclear data using multigroup nuclear data covariances. To verify the extension, sensitivity profiles of nuclear reactions are calculated via direct perturbation for the benchmark problems Jezebel, Godiva, LEU-SOL-THERM-002. The sensitivity profiles agree well with those obtained from TSUNAMI and Serpent. Secondly, the extension to XSUSA is applied to produce randomly sampled continuous-energy data libraries using the covariance libraries of SCALE 6.2. With these data libraries, samples of Serpent calculations are performed for Jezebel, Godiva, LEU-SOL-THERM-002, and the TMI-1 pin cell of the OECD/NEA LWR-UAM benchmark. For each problem, the multiplication factor uncertainty agrees well with the one from TSUNAMI.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110955"},"PeriodicalIF":1.9,"publicationDate":"2024-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422948","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-07DOI: 10.1016/j.anucene.2024.110961
Cyprien Richard , Mathias François , Lucas Fede , Alain Hébert
Open source modeling of VVER-type reactors could become a medium-term objective in Eastern Europe. As the deterministic code DRAGON5 could meet such a need, we confronted DRAGON5 against a stochastic reference code, SERPENT2. Our validation comprises 7 cells and 4 assemblies from the Khmelnitsky-2 reactor in Ukraine, within a wide range of heterogeneity levels in fuel composition. Two calculation schemes have been developed and compared. The first, the ALAMOS scheme, is highly discretized in energy and spatial resolution, while the second, the REL2005-like scheme, is calculated in two levels (one highly discretized in energy and the other highly discretized in space). In the majority of cases studied, both schemes offer satisfactory accuracy (e.g. less than 300 pcm in ), although there are difficulties related to energy deposition with gadolinium-poisoned fuel. While showing significantly poorer results than the ALAMOS scheme, the REL2005-like scheme offers lower computation times and major avenues for improvement remain to be explored. This work offers a first step towards the simulation of VVER-type reactors in DRAGON5, and paves the way for full-core simulations.
{"title":"Development of a computational scheme based on the DRAGON5 code for the neutronic study of VVER-type reactor rods and assemblies","authors":"Cyprien Richard , Mathias François , Lucas Fede , Alain Hébert","doi":"10.1016/j.anucene.2024.110961","DOIUrl":"10.1016/j.anucene.2024.110961","url":null,"abstract":"<div><div>Open source modeling of VVER-type reactors could become a medium-term objective in Eastern Europe. As the deterministic code DRAGON5 could meet such a need, we confronted DRAGON5 against a stochastic reference code, SERPENT2. Our validation comprises 7 cells and 4 assemblies from the Khmelnitsky-2 reactor in Ukraine, within a wide range of heterogeneity levels in fuel composition. Two calculation schemes have been developed and compared. The first, the ALAMOS scheme, is highly discretized in energy and spatial resolution, while the second, the REL2005-like scheme, is calculated in two levels (one highly discretized in energy and the other highly discretized in space). In the majority of cases studied, both schemes offer satisfactory accuracy (e.g. less than 300 pcm in <span><math><msub><mrow><mi>k</mi></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>), although there are difficulties related to energy deposition with gadolinium-poisoned fuel. While showing significantly poorer results than the ALAMOS scheme, the REL2005-like scheme offers lower computation times and major avenues for improvement remain to be explored. This work offers a first step towards the simulation of VVER-type reactors in DRAGON5, and paves the way for full-core simulations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110961"},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422987","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-07DOI: 10.1016/j.anucene.2024.110956
Fanli Kong, Xu Cheng
In severe accidents of nuclear power plants, large amounts of fission products existing as radioactive aerosols are released. Pool scrubbing plays an important role in the removal of radioactive aerosols. Bubble residence time is one of the key parameters to determine the efficiency of aerosol removal, especially in the swarm flow region which makes a very important contribution to the total aerosol removal. In this study, the Euler-Euler-Lagrangian approach is built to track the evolution of bubble motion and to determine the bubble residence time in the liquid pool. Specifically, the Euler-Euler two-fluid approach is utilized to resolve the flow field of gas and liquid phases, while the Lagrangian approach is employed to track the discrete bubbles and to obtain the bubble residence time. The results reveal that the present approach is feasible to predict the bubble dynamics and residence time in the liquid pool. Bubble residence time is dependent on the initial position, where bubbles deviating from the central region could remain inside the liquid pool for a longer physical time. The bubble diameter, volume flow rate and submergence height are key parameters affecting the bubble residence time. And comparison between the simulated bubble residence time and the model-predicted results is carried out, indicating the discrepancy of simulated residence time and limitations of the existing model at high volume flow rate and high submergence.
{"title":"Analysis of swarm flow and bubble residence time under pool scrubbing conditions","authors":"Fanli Kong, Xu Cheng","doi":"10.1016/j.anucene.2024.110956","DOIUrl":"10.1016/j.anucene.2024.110956","url":null,"abstract":"<div><div>In severe accidents of nuclear power plants, large amounts of fission products existing as radioactive aerosols are released. Pool scrubbing plays an important role in the removal of radioactive aerosols. Bubble residence time is one of the key parameters to determine the efficiency of aerosol removal, especially in the swarm flow region which makes a very important contribution to the total aerosol removal. In this study, the Euler-Euler-Lagrangian approach is built to track the evolution of bubble motion and to determine the bubble residence time in the liquid pool. Specifically, the Euler-Euler two-fluid approach is utilized to resolve the flow field of gas and liquid phases, while the Lagrangian approach is employed to track the discrete bubbles and to obtain the bubble residence time. The results reveal that the present approach is feasible to predict the bubble dynamics and residence time in the liquid pool. Bubble residence time is dependent on the initial position, where bubbles deviating from the central region could remain inside the liquid pool for a longer physical time. The bubble diameter, volume flow rate and submergence height are key parameters affecting the bubble residence time. And comparison between the simulated bubble residence time and the model-predicted results is carried out, indicating the discrepancy of simulated residence time and limitations of the existing model at high volume flow rate and high submergence.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110956"},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422945","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-07DOI: 10.1016/j.anucene.2024.110951
Shohanul Islam, Md Tanvir Ahmed
This study investigates the neutronics characterization of the Allegro-75 MWth modular reactor by analyzing three alternative heterogeneous core configurations-axial, radial, axial + radial across three fuel candidates-UPuC, UPuN, and UPuO along with reflector materials namely ZrC, SiC, BeO, and Zr3Sc2 to improve neutronics performance, identify the most suitable core configuration and optimal axial reflector thickness. The study revealed that axial + radial heterogeneous core configuration exhibited better performance across each fuel type compared to other heterogeneous models. UPuC with axial + radial heterogeneity was identified as the optimal model as it demonstrated cycle length over ten years, satisfactory neutron spectrum, uniform neutron flux distribution, low radial and axial PPF, high beta effective, and negative Doppler Constant. Analyzing reflector materials with the most suitable fuel model revealed that the optimum axial reflector thickness is 60 cm for all reflector models where BeO emerged as the most favorable reflector due to its superior results in other neutronics parameters.
{"title":"Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor","authors":"Shohanul Islam, Md Tanvir Ahmed","doi":"10.1016/j.anucene.2024.110951","DOIUrl":"10.1016/j.anucene.2024.110951","url":null,"abstract":"<div><div>This study investigates the neutronics characterization of the Allegro-75 MW<sub>th</sub> modular reactor by analyzing three alternative heterogeneous core configurations-axial, radial, axial + radial across three fuel candidates-UPuC, UPuN, and UPuO along with reflector materials namely ZrC, SiC, BeO, and Zr<sub>3</sub>Sc<sub>2</sub> to improve neutronics performance, identify the most suitable core configuration and optimal axial reflector thickness. The study revealed that axial + radial heterogeneous core configuration exhibited better performance across each fuel type compared to other heterogeneous models. UPuC with axial + radial heterogeneity was identified as the optimal model as it demonstrated cycle length over ten years, satisfactory neutron spectrum, uniform neutron flux distribution, low radial and axial PPF, high beta effective, and negative Doppler Constant. Analyzing reflector materials with the most suitable fuel model revealed that the optimum axial reflector thickness is 60 cm for all reflector models where BeO emerged as the most favorable reflector due to its superior results in other neutronics parameters.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110951"},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422944","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-05DOI: 10.1016/j.anucene.2024.110945
Kyungmin Kim, Minseung Ko, Sangtae Kim, Yongsoo Kim
Although the clearance level of every radioactive nuclide was published by the IAEA to promote the recycling and reuse of decontaminated radioactive waste worldwide, technical and regulatory issues have always been raised around the application of the criteria. Therefore, several countries are developing in-situ characterization equipment or apparatus for on-site verification to check if the clearance criteria is met.
In this study authors developed a pilot radiation detection and measurement system using in-situ characterization technology to solve the issues, which consists of a 3D scanning camera system and a built-in Monte Carlo simulation program. Measurement results show that MDA (Minimum Detectable Activity) of the current design was indisputably below the clearance level and built-in Monte Carlo simulation package closely predicts the measurements results with the error of less than 5%. This implicates that it can determine with enough margin whether the radioactivity level of decontaminated metallic components meets the clearance criteria at decommissioning site or not.
Practically when we measure the radioactivity from gamma ray source mass attenuation always takes place during the photon transports through the medium. In fact, the reduction depends on the material, shapes, and radioactive sources. In this study the reduction factors were experimentally examined according to the influencing parameters and the results were saved as DCF (Density Correction Factor) in the data base. As expected, it turned out that the factor is somewhat affected by medium material and radioactive sources, but it is basically proportional to the distance of gamma ray passage.
It is expected that upgraded design with more accurate and reliable instruments can make it easier for regulators to accept the application of the in-situ characterization technology on-site.
{"title":"A study on in-situ characterization technology development for clearance verification of radioactive waste from nuclear decommissioning","authors":"Kyungmin Kim, Minseung Ko, Sangtae Kim, Yongsoo Kim","doi":"10.1016/j.anucene.2024.110945","DOIUrl":"10.1016/j.anucene.2024.110945","url":null,"abstract":"<div><div>Although the clearance level of every radioactive nuclide was published by the IAEA to promote the recycling and reuse of decontaminated radioactive waste worldwide, technical and regulatory issues have always been raised around the application of the criteria. Therefore, several countries are developing in-situ characterization equipment or apparatus for on-site verification to check if the clearance criteria is met.</div><div>In this study authors developed a pilot radiation detection and measurement system using in-situ characterization technology to solve the issues, which consists of a 3D scanning camera system and a built-in Monte Carlo simulation program. Measurement results show that MDA (Minimum Detectable Activity) of the current design was <span><span>indisputably</span><svg><path></path></svg></span> below the clearance level and built-in Monte Carlo simulation package closely predicts the measurements results with the error of less than 5%. This implicates that it can determine with enough margin whether the radioactivity level of decontaminated metallic components meets the clearance criteria at decommissioning site or not.</div><div>Practically when we measure the radioactivity from gamma ray source mass attenuation always takes place during the photon transports through the medium. In fact, the reduction depends on the material, shapes, and radioactive sources. In this study the reduction factors were experimentally examined according to the influencing parameters and the results were saved as DCF (Density Correction Factor) in the data base. As expected, it turned out that the factor is somewhat affected by medium material and radioactive sources, but it is basically proportional to the distance of gamma ray passage.</div><div>It is expected that upgraded design with more accurate and reliable instruments can make it easier for regulators to accept the application of the in-situ characterization technology on-site.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110945"},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422991","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-05DOI: 10.1016/j.anucene.2024.110958
Mohamed A.E.M. Ali , Mohammed A.Y. Hafez , Nabil M. Nagy , Neveen S. Abed
In concrete applications. Major/critical applications of such concrete are radiation-shielding facilities. Both steel slag and silica fume are examples of common by-product materials that can be used as a replacer of aggregates and cement. Thus, in this research work, steel slag was utilized as heavy aggregate in concrete production besides silica fume to present sustainable concrete mixtures probably with better radiation-shielding properties. Different cementitious plasters were applied on the conducted sustainable concrete mixture using different powdery materials; hematite, magnetite, barite, bentonite, and steel slag powders in addition to nano-titanium dioxide as full replacers for sand. The proposed plasters were presented to determine the optimum plaster technique in terms of static performance and attenuation capability against gamma and neutron radiations. The results exhibited that utilizing steel slag and silica fume in concrete mixtures enhanced compressive strength by up to 9.09 % compared to conventional concrete, while the addition of nano-titanium to conventional plaster led to superior enhancement in the compressive strength by up to 38.65 % relative to traditional plaster. Conversely, fully replacing conventional silica sand with the abovementioned powdery materials generally reduced the compressive strength of cementitious plasters by up to 30.83 %. However, the radiation shielding properties against Cs-137, and Co-60 energies have been enhanced by up to 20 % and 26 %, respectively.
{"title":"Radiation shielding properties of sustainable concrete with novel plastering techniques","authors":"Mohamed A.E.M. Ali , Mohammed A.Y. Hafez , Nabil M. Nagy , Neveen S. Abed","doi":"10.1016/j.anucene.2024.110958","DOIUrl":"10.1016/j.anucene.2024.110958","url":null,"abstract":"<div><div>In concrete applications. Major/critical applications of such concrete are radiation-shielding facilities. Both steel slag and silica fume are examples of common by-product materials that can be used as a replacer of aggregates and cement. Thus, in this research work, steel slag was utilized as heavy aggregate in concrete production besides silica fume to present sustainable concrete mixtures probably with better radiation-shielding properties. Different cementitious plasters were applied on the conducted sustainable concrete mixture using different powdery materials; hematite, magnetite, barite, bentonite, and steel slag powders in addition to nano-titanium dioxide as full replacers for sand. The proposed plasters were presented to determine the optimum plaster technique in terms of static performance and attenuation capability against gamma and neutron radiations. The results exhibited that utilizing steel slag and silica fume in concrete mixtures enhanced compressive strength by up to 9.09 % compared to conventional concrete, while the addition of nano-titanium to conventional plaster led to superior enhancement in the compressive strength by up to 38.65 % relative to traditional plaster. Conversely, fully replacing conventional silica sand with the abovementioned powdery materials generally reduced the compressive strength of cementitious plasters by up to 30.83 %. However, the radiation shielding properties against Cs-137, and Co-60 energies have been enhanced by up to 20 % and 26 %, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110958"},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422942","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-05DOI: 10.1016/j.anucene.2024.110930
Michał Jędrzejczyk , Piotr Kopka , Basma Foad
Bayesian updating (BU) tools are increasingly used in nuclear engineering for inverse uncertainty quantification (IUQ) and calibration combined with uncertainty reduction. Their goals are quantifying or reducing the uncertainty of some of the uncertain model input parameters. Since it is often the case that available experimental data only allows for updating the most influential input parameters, researchers often ignore the less important ones during BU. This paper explores the consequences of neglecting the uncertainties of uncalibrated model input parameters (UMIP). It also proposes how to include them properly and which BU algorithms are the best choices for various types of inverse problems. The analysis is based on exploring two toy problems and one in nuclear engineering concerning multiplication factor calculations. The results clearly show that the improper treatment of UMIP during BU often leads to underestimating posterior uncertainties — either of the calibrated input parameters or the simulated integral parameters, depending on how the BU was conducted. The proposed methods of correct UMIP treatment will improve the rigorousness of the BU processes and boost confidence in the resulting posterior distributions.
贝叶斯更新（BU）工具越来越多地用于核工程中的反向不确定性量化（IUQ）和校准以及不确定性降低。其目标是量化或减少某些不确定模型输入参数的不确定性。由于可用的实验数据通常只允许更新影响最大的输入参数，因此研究人员在进行 BU 时往往会忽略不太重要的参数。本文探讨了忽略未校准模型输入参数（UMIP）不确定性的后果。本文还提出了如何正确地包含这些不确定性，以及哪些 BU 算法是各类逆问题的最佳选择。分析基于对两个玩具问题和一个核工程中的乘法因子计算问题的探讨。结果清楚地表明，BU 期间对 UMIP 的不当处理往往会导致低估后验不确定性--无论是校准输入参数还是模拟积分参数的后验不确定性，这取决于 BU 是如何进行的。所提出的正确处理 UMIP 的方法将提高 BU 过程的严谨性，并增强对所得后验分布的信心。
{"title":"Framework for the correct treatment of model input parameters for Bayesian updating problems in nuclear engineering","authors":"Michał Jędrzejczyk , Piotr Kopka , Basma Foad","doi":"10.1016/j.anucene.2024.110930","DOIUrl":"10.1016/j.anucene.2024.110930","url":null,"abstract":"<div><div>Bayesian updating (BU) tools are increasingly used in nuclear engineering for inverse uncertainty quantification (IUQ) and calibration combined with uncertainty reduction. Their goals are quantifying or reducing the uncertainty of some of the uncertain model input parameters. Since it is often the case that available experimental data only allows for updating the most influential input parameters, researchers often ignore the less important ones during BU. This paper explores the consequences of neglecting the uncertainties of uncalibrated model input parameters (UMIP). It also proposes how to include them properly and which BU algorithms are the best choices for various types of inverse problems. The analysis is based on exploring two toy problems and one in nuclear engineering concerning multiplication factor calculations. The results clearly show that the improper treatment of UMIP during BU often leads to underestimating posterior uncertainties — either of the calibrated input parameters or the simulated integral parameters, depending on how the BU was conducted. The proposed methods of correct UMIP treatment will improve the rigorousness of the BU processes and boost confidence in the resulting posterior distributions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110930"},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422943","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-04DOI: 10.1016/j.anucene.2024.110960
Donny Hartanto, David Chandler, Hailey Green, Jin Whan Bae, Kevin M. Burg, Yves Robert, Carol Sizemore
This paper presents a series of 252Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because 252Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce 252Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. The findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares 252Cf production and safety metrics with the HEU core and a proposed LEU core.
{"title":"Californium-252 production at the High Flux Isotope Reactor − I: Validation study using campaign data","authors":"Donny Hartanto, David Chandler, Hailey Green, Jin Whan Bae, Kevin M. Burg, Yves Robert, Carol Sizemore","doi":"10.1016/j.anucene.2024.110960","DOIUrl":"10.1016/j.anucene.2024.110960","url":null,"abstract":"<div><div>This paper presents a series of <sup>252</sup>Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because <sup>252</sup>Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce <sup>252</sup>Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. The findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares <sup>252</sup>Cf production and safety metrics with the HEU core and a proposed LEU core.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110960"},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422990","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In recent years, nuclear forensic analysis has become crucial due to the growing global threat of nuclear terrorism and smuggling. Since 2005, extensive research has been conducted on identifying the origin of spent nuclear fuel, focusing on the source reactor-type discrimination, 235U enrichment of the fresh fuel, and the fuel exposure in the reactor (known as burnup). However, the majority of research relies on computed databases, which may lead to tracing discrepancies compared with actual situations. The present study employs the isotopic measurements from the experimental SFCOMPO-2.0 database to predict nuclear reactor types using Factor Analysis (FA) and various machine learning classification algorithms. The results reveal that FA is an effective method for dimension reduction and visualization. The FA-KNN, Random Forest (RF), and Multilayer Perceptron (MLP) algorithms are applied using a consistent dataset partition to ensure unbiased comparisons. The prediction results based on 10-fold stratified cross-validation are quite promising and the Receiver Operating Characteristic (ROC) curves for multi-class classification confirm the excellent generalization ability of models. Therefore, the application of machine learning techniques is highly effective for reactor-type forensics analysis, especially for RF and MLP.
{"title":"Advancing source reactor-type discrimination using machine learning techniques and SFCOMPO-2.0 experimental database","authors":"Tianxiang Wang, Hao Yang, Shengli Chen, Cenxi Yuan","doi":"10.1016/j.anucene.2024.110952","DOIUrl":"10.1016/j.anucene.2024.110952","url":null,"abstract":"<div><div>In recent years, nuclear forensic analysis has become crucial due to the growing global threat of nuclear terrorism and smuggling. Since 2005, extensive research has been conducted on identifying the origin of spent nuclear fuel, focusing on the source reactor-type discrimination, <sup>235</sup>U enrichment of the fresh fuel, and the fuel exposure in the reactor (known as burnup). However, the majority of research relies on computed databases, which may lead to tracing discrepancies compared with actual situations. The present study employs the isotopic measurements from the experimental SFCOMPO-2.0 database to predict nuclear reactor types using Factor Analysis (FA) and various machine learning classification algorithms. The results reveal that FA is an effective method for dimension reduction and visualization. The FA-KNN, Random Forest (RF), and Multilayer Perceptron (MLP) algorithms are applied using a consistent dataset partition to ensure unbiased comparisons. The prediction results based on 10-fold stratified cross-validation are quite promising and the Receiver Operating Characteristic (ROC) curves for multi-class classification confirm the excellent generalization ability of models. Therefore, the application of machine learning techniques is highly effective for reactor-type forensics analysis, especially for RF and MLP.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110952"},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142422989","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-03DOI: 10.1016/j.anucene.2024.110937
O. Halim , F. Galleni , N. Forgione , I. Di Piazza , A. Pucciarelli
The paper investigates the capabilities of different CFD modelling approaches in reproducing operating conditions relevant for Liquid Metal Fast Breeder Reactors technologies. The selected benchmark is the NACIE-UP facility wire-wrapped fuel bundle using Lead-Bismuth Eutectic (LBE) as coolant: the predictions are compared to the experimental data collected for several operating conditions considered in the frame of two distinct experimental campaigns. Four different modelling approaches have been adopted in this work to model the NACIE-UP Fuel Pin Simulator: Bare, Detailed, Solid-Wire and the Porous-Wire Rod Bundle model. A model-to-model comparison is performed to understand the benefits, limitations, and accuracy of using different modelling approaches for representing wrapped wires fuel bundles. Furthermore, integrating NACIE-UP benchmark experimental data into the comparative analysis reinforce the validation process of the adopted modelling approaches.
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