Pub Date : 2024-10-28DOI: 10.1016/j.anucene.2024.111003
Yuan Chang , Hui Wang , Gong-Lin Li , Ming Ding
An in-house code called CONTHAC-3D was developed to investigate the fundamental thermal–hydraulic phenomena occurred in the containment under severe accidents for NPPs. The code included specific models to simulate the special systems of HPR1000 and ACP100. The classical backward-facing step flow benchmark and BMC HYJET helium jet experiments were selected to investigate the code’s capability of simulating hydrogen diffusion process. The results showed that the difference between the calculated and experimental results could be negligible. The code was then applied to investigate hydrogen diffusion and distribution for HPR1000. The results showed that the hydrogen released from the break rises vertically and rapidly to the containment dome, then the gas diffused into the dome and lower compartments. As the time went by, the hydrogen concentration in lower compartments seemed to be higher than that in the containment dome. The results could provide foundation for the arrangement of hydrogen risk mitigation measures.
{"title":"Research on simulation of hydrogen diffusion behavior based on CONTHAC-3D code","authors":"Yuan Chang , Hui Wang , Gong-Lin Li , Ming Ding","doi":"10.1016/j.anucene.2024.111003","DOIUrl":"10.1016/j.anucene.2024.111003","url":null,"abstract":"<div><div>An in-house code called CONTHAC-3D was developed to investigate the fundamental thermal–hydraulic phenomena occurred in the containment under severe accidents for NPPs. The code included specific models to simulate the special systems of HPR1000 and ACP100. The classical backward-facing step flow benchmark and BMC HYJET helium jet experiments were selected to investigate the code’s capability of simulating hydrogen diffusion process. The results showed that the difference between the calculated and experimental results could be negligible. The code was then applied to investigate hydrogen diffusion and distribution for HPR1000. The results showed that the hydrogen released from the break rises vertically and rapidly to the containment dome, then the gas diffused into the dome and lower compartments. As the time went by, the hydrogen concentration in lower compartments seemed to be higher than that in the containment dome. The results could provide foundation for the arrangement of hydrogen risk mitigation measures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111003"},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527102","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 : 2024-10-28DOI: 10.1016/j.anucene.2024.110978
J.L. Wormald, A.J. Trainer, M.L. Zerkle
A new method is presented to use machine-learned interatomic potentials (MLPs) to generate material models for thermal neutron scattering laws (TSLs). MLPs are computationally efficient models of ab initio force fields that can be used in the creation of a vibrational spectrum as an input to TSL generation. MLP-based molecular dynamics introduces temperature effects into the vibrational spectrum, which have been neglected in most modern TSLs. Yttrium hydride () is used to illustrate this new MLP technique. The MLP approach is shown to predict temperature effects in the vibrational spectrum observed in experiment and improve on key features of the oscillatory scattering cross section of when compared to current temperature-independent, ab initio techniques.
{"title":"Machine-learned force fields for thermal neutron scattering law evaluations","authors":"J.L. Wormald, A.J. Trainer, M.L. Zerkle","doi":"10.1016/j.anucene.2024.110978","DOIUrl":"10.1016/j.anucene.2024.110978","url":null,"abstract":"<div><div>A new method is presented to use machine-learned interatomic potentials (MLPs) to generate material models for thermal neutron scattering laws (TSLs). MLPs are computationally efficient models of <em>ab initio</em> force fields that can be used in the creation of a vibrational spectrum as an input to TSL generation. MLP-based molecular dynamics introduces temperature effects into the vibrational spectrum, which have been neglected in most modern TSLs. Yttrium hydride (<span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span>) is used to illustrate this new MLP technique. The MLP approach is shown to predict temperature effects in the vibrational spectrum observed in experiment and improve on key features of the oscillatory scattering cross section of <span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span> when compared to current temperature-independent, <em>ab initio</em> techniques.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110978"},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527101","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 : 2024-10-26DOI: 10.1016/j.anucene.2024.111017
Jaehong Lee , Fabiana Rossi , Yu Kodama , Kota Hironaka , Mitsuo Koizumi , Tadafumi Sano , Yasunori Matsuo , Jun-ichi Hori
Silica glass has been used as a base and host material in vitrified radioactive waste and lithium glass scintillators for neutron detection because of its superb transparency, high heat resistance, and excellent chemical inertness. Therefore, an accurate total cross section of the silica glass is crucial to evaluate the criticality safety of vitrified wastes and understand the neutron response for lithium glass scintillators. This study performed neutron transmission measurements for silica glass using a pulsed neutron beam with the time-of-flight method at the Kyoto University Institute for Integrated Radiation and Nuclear Science − Linear Accelerator to provide an accurate total cross section in the thermal and epithermal energy range. We obtained the neutron total cross section of the silica glass in the energy region from 0.002–25 eV. The results were compared and discussed with previous results and evaluated data.
硅玻璃因其极佳的透明度、高耐热性和优异的化学惰性，一直被用作玻璃化放射性废物和中子探测锂玻璃闪烁体的基体和主材料。因此，精确的硅玻璃总截面对于评估玻璃化废物的临界安全性和了解锂玻璃闪烁体的中子响应至关重要。本研究在京都大学综合辐射与核科学研究所--直线加速器使用脉冲中子束和飞行时间法对硅玻璃进行了中子透射测量，以提供热能和表热能范围内的精确总截面。我们获得了二氧化硅玻璃在 0.002-25 eV 能量范围内的中子总截面。我们将结果与以前的结果和评估数据进行了比较和讨论。
{"title":"Neutron transmission measurements for silica glass at the KURNS-LINAC","authors":"Jaehong Lee , Fabiana Rossi , Yu Kodama , Kota Hironaka , Mitsuo Koizumi , Tadafumi Sano , Yasunori Matsuo , Jun-ichi Hori","doi":"10.1016/j.anucene.2024.111017","DOIUrl":"10.1016/j.anucene.2024.111017","url":null,"abstract":"<div><div>Silica glass has been used as a base and host material in vitrified radioactive waste and lithium glass scintillators for neutron detection because of its superb transparency, high heat resistance, and excellent chemical inertness. Therefore, an accurate total cross section of the silica glass is crucial to evaluate the criticality safety of vitrified wastes and understand the neutron response for lithium glass scintillators. This study performed neutron transmission measurements for silica glass using a pulsed neutron beam with the time-of-flight method at the Kyoto University Institute for Integrated Radiation and Nuclear Science − Linear Accelerator to provide an accurate total cross section in the thermal and epithermal energy range. We obtained the neutron total cross section of the silica glass in the energy region from 0.002–25 eV. The results were compared and discussed with previous results and evaluated data.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111017"},"PeriodicalIF":1.9,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527011","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 : 2024-10-25DOI: 10.1016/j.anucene.2024.110988
Előd Pázmán , Gábor Tolnai , Dávid Légrády , Luigi Mercatali , Gianfranco Huaccho , Victor Hugo Sanchez-Espinoza
GUARDYAN is a dynamic 3D Monte Carlo reactor physics code with continuous energy handling developed for GPU hardware that has recently been coupled to the SUBCHANFLOW (SCF) subchannel thermal hydraulics solver. In this paper two control rod ejection accident scenarios will be presented in a Small Modular Reactor (SMR) geometry: a transient starting from Hot Zero Power (HZP), and one starting from Hot Full Power (HFP) conditions, both of them using Beginning of Cycle (BOC) material composition. Both the time dependent core-wise data and the node-wise data at certain times calculated by the GUARDYAN-SCF coupled code system exhibit the tendencies expected during such transients, with the thermal hydraulic properties mostly inside their safe limits. Relative variances estimated from 8 independent realisations suggest the results are credible. To further support our findings the HZP results are presented alongside data from PARCS-SCF and Serpent2-SCF calculations provided by Karlsruhe Institute of Technology (KIT), while for the HFP case we were able to compare some of the quantities to PARCS-SCF results.
{"title":"HZP and HFP rod ejection analysis in a SMART-like reactor model using the GUARDYAN-SUBCHANFLOW coupled code system","authors":"Előd Pázmán , Gábor Tolnai , Dávid Légrády , Luigi Mercatali , Gianfranco Huaccho , Victor Hugo Sanchez-Espinoza","doi":"10.1016/j.anucene.2024.110988","DOIUrl":"10.1016/j.anucene.2024.110988","url":null,"abstract":"<div><div>GUARDYAN is a dynamic 3D Monte Carlo reactor physics code with continuous energy handling developed for GPU hardware that has recently been coupled to the SUBCHANFLOW (SCF) subchannel thermal hydraulics solver. In this paper two control rod ejection accident scenarios will be presented in a Small Modular Reactor (SMR) geometry: a transient starting from Hot Zero Power (HZP), and one starting from Hot Full Power (HFP) conditions, both of them using Beginning of Cycle (BOC) material composition. Both the time dependent core-wise data and the node-wise data at certain times calculated by the GUARDYAN-SCF coupled code system exhibit the tendencies expected during such transients, with the thermal hydraulic properties mostly inside their safe limits. Relative variances estimated from 8 independent realisations suggest the results are credible. To further support our findings the HZP results are presented alongside data from PARCS-SCF and Serpent2-SCF calculations provided by Karlsruhe Institute of Technology (KIT), while for the HFP case we were able to compare some of the quantities to PARCS-SCF results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110988"},"PeriodicalIF":1.9,"publicationDate":"2024-10-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527013","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 : 2024-10-24DOI: 10.1016/j.anucene.2024.110999
Tomáš Czakoj , Michal Košťál , Evžen Novák , Jan Šimon , Martin Schulc , Zdeněk Matěj , Filip Mravec , František Cvachovec , Tomáš Urban , Evžen Losa
The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures.
A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, a marked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.
{"title":"Measurement of gamma field inside the biological concrete shielding of VVER-1000 Mock-Up at the LR-0 reactor","authors":"Tomáš Czakoj , Michal Košťál , Evžen Novák , Jan Šimon , Martin Schulc , Zdeněk Matěj , Filip Mravec , František Cvachovec , Tomáš Urban , Evžen Losa","doi":"10.1016/j.anucene.2024.110999","DOIUrl":"10.1016/j.anucene.2024.110999","url":null,"abstract":"<div><div>The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures.</div><div>A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, a<!--> <!-->marked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110999"},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526955","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 : 2024-10-24DOI: 10.1016/j.anucene.2024.110998
Sohail Ahmad Raza, Muhammad Hashim, Liangzhi Cao, Xianan Du, Longwen Jiang
SPARK-NC, a 10 MW(e) lead–bismuth eutectic (LBE) cooled fast reactor design, exhibits promising characteristics like inherent gamma shielding, natural circulation, and a high boiling point. Following detailed neutronic studies, a thorough investigation of nuclear safety necessitates a detailed analysis of the core radionuclide inventory. This information is particularly crucial for source term calculations, which play a vital role in assessing the potential radiological consequences. This study establishes the life-cycle inventory of SPARK-NC using two independent computational systems: ORIGEN2.2 and NECP-SARAX. ORIGEN2.2, equipped with a reactor-specific library generated by NECP-MCX, is used for average whole-core inventory analysis. NECP-SARAX, on the other hand, explicitly considers core heterogeneity in terms of enrichment, specific power, and burn-up. This work presents the radionuclide inventories and the relative calculation differences observed between the codes. Actinides like uranium and curium display minimal code dependence, while plutonium isotopes exhibit a maximum relative difference of 8 %. Fission products generally agree within 5 %, except for I-131, which shows a discrepancy of around 10 %. The activity of I-131 and Cs-137 are estimated to be approximately 1 × 1016 Bq and 3 × 1015 Bq, respectively. Additionally, the photon source strength is 1017/s at 1 MeV, dropping to 1016/s below 6 MeV. Fission products and actinides contribute a decay heat of 0.65 MW. Assembly-wise analysis reveals a direct proportionality between radionuclide inventory and peaking factor, with the average assembly inventory being roughly 25 % lower than the peak assembly inventory. Rare earth elements (Ce, Sm, Pm, Pr, Nd, La, Y) exhibit a maximum mass of approximately 8.5 kg with a 3 % relative difference between the codes. Conversely, halogens (I, Br) have a minimum mass of around 0.2 kg with a 13 % relative difference. These findings, alongside the quantification of radionuclides, provide valuable insights into the code selection for future analyses of SPARK-NC and similar reactor systems.
{"title":"Investigation on radioisotopes evolution in the fuel of Lead-Bismuth eutectic (LBE) cooled SPARK-NC core","authors":"Sohail Ahmad Raza, Muhammad Hashim, Liangzhi Cao, Xianan Du, Longwen Jiang","doi":"10.1016/j.anucene.2024.110998","DOIUrl":"10.1016/j.anucene.2024.110998","url":null,"abstract":"<div><div>SPARK-NC, a 10 MW(e) lead–bismuth eutectic (LBE) cooled fast reactor design, exhibits promising characteristics like inherent gamma shielding, natural circulation, and a high boiling point. Following detailed neutronic studies, a thorough investigation of nuclear safety necessitates a detailed analysis of the core radionuclide inventory. This information is particularly crucial for source term calculations, which play a vital role in assessing the potential radiological consequences. This study establishes the life-cycle inventory of SPARK-NC using two independent computational systems: ORIGEN2.2 and NECP-SARAX. ORIGEN2.2, equipped with a reactor-specific library generated by NECP-MCX, is used for average whole-core inventory analysis. NECP-SARAX, on the other hand, explicitly considers core heterogeneity in terms of enrichment, specific power, and burn-up. This work presents the radionuclide inventories and the relative calculation differences observed between the codes. Actinides like uranium and curium display minimal code dependence, while plutonium isotopes exhibit a maximum relative difference of 8 %. Fission products generally agree within 5 %, except for I-131, which shows a discrepancy of around 10 %. The activity of I-131 and Cs-137 are estimated to be approximately 1 × 10<sup>16</sup> Bq and 3 × 10<sup>15</sup> Bq, respectively. Additionally, the photon source strength is 10<sup>17</sup>/s at 1 MeV, dropping to 10<sup>16</sup>/s below 6 MeV. Fission products and actinides contribute a decay heat of 0.65 MW. Assembly-wise analysis reveals a direct proportionality between radionuclide inventory and peaking factor, with the average assembly inventory being roughly 25 % lower than the peak assembly inventory. Rare earth elements (Ce, Sm, Pm, Pr, Nd, La, Y) exhibit a maximum mass of approximately 8.5 kg with a 3 % relative difference between the codes. Conversely, halogens (I, Br) have a minimum mass of around 0.2 kg with a 13 % relative difference. These findings, alongside the quantification of radionuclides, provide valuable insights into the code selection for future analyses of SPARK-NC and similar reactor systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110998"},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526931","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 : 2024-10-24DOI: 10.1016/j.anucene.2024.110966
Hikaru Hiruta, Mark D. DeHart, Carlo Parisi
This work investigates the option of modifying the Advanced Test Reactor by replacing the current beryllium reflector with heavy water. Such a change may provide some potential benefits for not only increasing the thermal irradiation capabilities but also resolving other problems such as reflector integrity issues due to fast fluence damage, which is always a limiting factor in the lifetime of the current beryllium reflector. This paper presents the analysis and estimation of the ATR core physics parameters by replacing the current beryllium reflector with heavy water (D2O). The paper first describes the details of two selected conceptual designs, which are partially reflected with either beryllium or graphite, and how they are derived from the baseline beryllium reflector concept. Then, reactor physics performance parameters for the two new concepts are assessed by comparing with those of the baseline concept. The performance parameters considered in this paper include in-pile tube neutron and gamma fluxes and heating rates, maximum loop voiding reactivity, core power behavior with different power splits, predicted cycle length with a given fuel loading, and thermal hydraulic analysis with a higher lobe power split. It is important to note that this study focuses on the reactor physics aspects and does not delve into the engineering challenges associated with such a design modification.
{"title":"Physics analysis and design of heavy water reflected thermal test reactor","authors":"Hikaru Hiruta, Mark D. DeHart, Carlo Parisi","doi":"10.1016/j.anucene.2024.110966","DOIUrl":"10.1016/j.anucene.2024.110966","url":null,"abstract":"<div><div>This work investigates the option of modifying the Advanced Test Reactor by replacing the current beryllium reflector with heavy water. Such a change may provide some potential benefits for not only increasing the thermal irradiation capabilities but also resolving other problems such as reflector integrity issues due to fast fluence damage, which is always a limiting factor in the lifetime of the current beryllium reflector. This paper presents the analysis and estimation of the ATR core physics parameters by replacing the current beryllium reflector with heavy water (D<sub>2</sub>O). The paper first describes the details of two selected conceptual designs, which are partially reflected with either beryllium or graphite, and how they are derived from the baseline beryllium reflector concept. Then, reactor physics performance parameters for the two new concepts are assessed by comparing with those of the baseline concept. The performance parameters considered in this paper include in-pile tube neutron and gamma fluxes and heating rates, maximum loop voiding reactivity, core power behavior with different power splits, predicted cycle length with a given fuel loading, and thermal hydraulic analysis with a higher lobe power split. It is important to note that this study focuses on the reactor physics aspects and does not delve into the engineering challenges associated with such a design modification.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110966"},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526956","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 : 2024-10-23DOI: 10.1016/j.anucene.2024.110928
Vincent Lamirand , Alexander Knospe , Klemen Ambrožič , Sebastian Hübner , Carsten Lange , Oskari Pakari , Fanny Vitullo , Adolfo Rais , Joachim Pohlus , Uwe Paquee , Christoph Pohl , Nicolas Weiss , Pavel Frajtag , Daniel Godat , Antonios Mylonakis , Axel Laureau , Thomas Ligonnet , Mathieu Hursin , Grégory Perret , Andreas Pautz
We present the experimental campaigns – namely, three per facility – carried out between 2018 and 2021 in the AKR–2 and CROCUS zero power reactors within the framework of the Horizon 2020 European project CORTEX. Their purpose was to produce high-quality and noise-specific experimental data for the validation of the neutron noise computational models developed in CORTEX. In both reactors, perturbations were induced by two devices, separately and altogether. In AKR–2, they consisted of a rotating absorber, i.e. an absorber of variable strength, and a linear oscillator, i.e. a vibrating absorber, both sets in horizontal channels close to the core. In CROCUS, the project benefited from the COLIBRI experimental program and its fuel rods oscillator set in the outer lattice; an additional vibrating absorber called POLLEN was set in a vertical air-channel at core center. The campaigns at both facilities consisted of neutron measurements with numerous detectors at reference static states, and with the addition of the mechanical perturbations to induce neutron reactivity modulation. The present article documents the experimental setups and measurements for each facility and perturbation type. A focus is set on the experimental designs and their evolution along the project, as well as motivations and learned lessons. Results are presented and discussed in details in associated papers.
{"title":"CORTEX experiments – Part I: Modulation campaigns in AKR-2 & CROCUS for the validation of neutron noise codes","authors":"Vincent Lamirand , Alexander Knospe , Klemen Ambrožič , Sebastian Hübner , Carsten Lange , Oskari Pakari , Fanny Vitullo , Adolfo Rais , Joachim Pohlus , Uwe Paquee , Christoph Pohl , Nicolas Weiss , Pavel Frajtag , Daniel Godat , Antonios Mylonakis , Axel Laureau , Thomas Ligonnet , Mathieu Hursin , Grégory Perret , Andreas Pautz","doi":"10.1016/j.anucene.2024.110928","DOIUrl":"10.1016/j.anucene.2024.110928","url":null,"abstract":"<div><div>We present the experimental campaigns –<!--> <!-->namely, three per facility<!--> <!-->– carried out between 2018 and 2021 in the AKR–2 and CROCUS zero power reactors within the framework of the Horizon 2020 European project CORTEX. Their purpose was to produce high-quality and noise-specific experimental data for the validation of the neutron noise computational models developed in CORTEX. In both reactors, perturbations were induced by two devices, separately and altogether. In AKR–2, they consisted of a rotating absorber, i.e. an <em>absorber of variable strength</em>, and a linear oscillator, i.e. a <em>vibrating absorber</em>, both sets in horizontal channels close to the core. In CROCUS, the project benefited from the COLIBRI experimental program and its <em>fuel rods oscillator</em> set in the outer lattice; an additional <em>vibrating absorber</em> called POLLEN was set in a vertical air-channel at core center. The campaigns at both facilities consisted of neutron measurements with numerous detectors at reference static states, and with the addition of the mechanical perturbations to induce neutron reactivity modulation. The present article documents the experimental setups and measurements for each facility and perturbation type. A focus is set on the experimental designs and their evolution along the project, as well as motivations and learned lessons. Results are presented and discussed in details in associated papers.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110928"},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526947","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 : 2024-10-23DOI: 10.1016/j.anucene.2024.110989
Jan Malec , Oscar Cabellos , Marjan Kromar , Andrej Trkov
The WIMSD-5B transport code is a deterministic tool for nuclear reactor core design and fuel management. It can efficiently handle pin-cell and supercell models and calculate homogenized cross sections, which are essential for reactor physics calculations. It is used by core design packages such as the CORD-2 package, developed at the Jožef Stefan Institute, and SEANAP developed by Universidad Politécnica de Madrid (UPM). The WLUP update project https://www-nds.iaea.org/wimsd demonstrated the way to update the WIMS-D libraries with different evaluated nuclear data libraries, including ENDF libraries up to version ENDF/B-VII.1. Using an updated version of the procedure, a new WIMS-D library based on the ENDF/B-VIII.0 data was developed to improve the accuracy of core design calculations. Several improvements to the library were made and the effects of each individual improvement was demonstrated using a 3×3 supercell benchmark model that is representative of a typical pressurized water reactor. Finally, the performance of the library over a diverse set of neutron transport problems was tested for, to ensure no regressions were introduced.
{"title":"Generation and validation of a new WIMS-D library based on ENDF/B-VIII.0","authors":"Jan Malec , Oscar Cabellos , Marjan Kromar , Andrej Trkov","doi":"10.1016/j.anucene.2024.110989","DOIUrl":"10.1016/j.anucene.2024.110989","url":null,"abstract":"<div><div>The WIMSD-5B transport code is a deterministic tool for nuclear reactor core design and fuel management. It can efficiently handle pin-cell and supercell models and calculate homogenized cross sections, which are essential for reactor physics calculations. It is used by core design packages such as the CORD-2 package, developed at the Jožef Stefan Institute, and SEANAP developed by Universidad Politécnica de Madrid (UPM). The WLUP update project <span><span>https://www-nds.iaea.org/wimsd</span><svg><path></path></svg></span> demonstrated the way to update the WIMS-D libraries with different evaluated nuclear data libraries, including ENDF libraries up to version ENDF/B-VII.1. Using an updated version of the procedure, a new WIMS-D library based on the ENDF/B-VIII.0 data was developed to improve the accuracy of core design calculations. Several improvements to the library were made and the effects of each individual improvement was demonstrated using a 3×3 supercell benchmark model that is representative of a typical pressurized water reactor. Finally, the performance of the library over a diverse set of neutron transport problems was tested for, to ensure no regressions were introduced.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110989"},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526952","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 : 2024-10-23DOI: 10.1016/j.anucene.2024.111007
Djahid Lababsa , Hakim Mazrou , Mohamed Belgaid , Tahar Zidi , Mohammed Azzoune , Azzeddine Ameur , Ahmed Guesmia , Leila Zamoun , Mohamed Boufenar
This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.
The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.
Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.
{"title":"Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence","authors":"Djahid Lababsa , Hakim Mazrou , Mohamed Belgaid , Tahar Zidi , Mohammed Azzoune , Azzeddine Ameur , Ahmed Guesmia , Leila Zamoun , Mohamed Boufenar","doi":"10.1016/j.anucene.2024.111007","DOIUrl":"10.1016/j.anucene.2024.111007","url":null,"abstract":"<div><div>This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.</div><div>The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.</div><div>Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111007"},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526948","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}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号