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}
Pub Date : 2024-10-22DOI: 10.1016/j.anucene.2024.110987
Zhang Dandi , Wang Shanpu , Tong Lili , Cao Xuewu
Aerosol retention inside narrow channels is the optimization direction of the leakage source term assessment for nuclear power plant containment. Based on the flow characteristics of carrier gas and the deposition characteristics of transported aerosol, a one-dimensional analysis method of aerosol retention in narrow channels is developed through considering different deposition mechanisms of inlet loss, gravity settlement, Brownian diffusion, turbulent deposition and steam condensation. The flow models of carrier gas and the retention models of aerosol are analyzed and verified, respectively. The flow of carrier gas deviates from laminar flow earlier through using the drag model of narrow channels. The prediction accuracy of aerosol penetration factor calculated by current analysis method in narrow channels is improved under laminar flow and turbulent flow through comparing with the previous calculation methods. Aerosol retention analysis is conducted on the narrow channels of steel containment under the typical severe accident. The turbulent deposition introduced by larger leakage channels increases the aerosols retention effect in narrow channels.
{"title":"Retention analysis of aerosol inside narrow channels of the containment","authors":"Zhang Dandi , Wang Shanpu , Tong Lili , Cao Xuewu","doi":"10.1016/j.anucene.2024.110987","DOIUrl":"10.1016/j.anucene.2024.110987","url":null,"abstract":"<div><div>Aerosol retention inside narrow channels is the optimization direction of the leakage source term assessment for nuclear power plant containment. Based on the flow characteristics of carrier gas and the deposition characteristics of transported aerosol, a one-dimensional analysis method of aerosol retention in narrow channels is developed through considering different deposition mechanisms of inlet loss, gravity settlement, Brownian diffusion, turbulent deposition and steam condensation. The flow models of carrier gas and the retention models of aerosol are analyzed and verified, respectively. The flow of carrier gas deviates from laminar flow earlier through using the drag model of narrow channels. The prediction accuracy of aerosol penetration factor calculated by current analysis method in narrow channels is improved under laminar flow and turbulent flow through comparing with the previous calculation methods. Aerosol retention analysis is conducted on the narrow channels of steel containment under the typical severe accident. The turbulent deposition introduced by larger leakage channels increases the aerosols retention effect in narrow channels.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110987"},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526850","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 : 2024-10-22DOI: 10.1016/j.anucene.2024.110986
Jingkang Li , Zunyan Hu , Zeguang Li , Liangfei Xu , Jianqiu Li
Heat pipe cooled reactors (HPRs) offer the potential to achieve load-following control without the need for control rods or drums, thereby simplifying the control system. However, during load-following operation, HPRs experience fluctuations in temperature, which can impact safety. Limited research has focused on mitigating temperature fluctuations of HPRs during dynamic power regulation leveraging their inherent load-following capabilities. This study examines the characteristics of an HPR with closed Brayton Cycle (CBC), and develops a load-following control algorithm. A simplified CBC model is proposed to facilitate control strategy analysis. Model predictive control (MPC) is employed to suppress temperature fluctuations, revealing that the dynamic response of output power under MPC resembles that of a first-order inertial system. Consequently, a power control algorithm based on first-order inertial feedforward control is introduced. Simulation results demonstrate that the proposed algorithm, with a time constant ranging between 500 and 1000 s, significantly mitigates temperature and power fluctuations in HPRs during load-following dynamic power regulation.
{"title":"Temperature fluctuation mitigation of heat pipe cooled reactor with closed Brayton cycle during load-following dynamic power regulation","authors":"Jingkang Li , Zunyan Hu , Zeguang Li , Liangfei Xu , Jianqiu Li","doi":"10.1016/j.anucene.2024.110986","DOIUrl":"10.1016/j.anucene.2024.110986","url":null,"abstract":"<div><div>Heat pipe cooled reactors (HPRs) offer the potential to achieve load-following control without the need for control rods or drums, thereby simplifying the control system. However, during load-following operation, HPRs experience fluctuations in temperature, which can impact safety. Limited research has focused on mitigating temperature fluctuations of HPRs during dynamic power regulation leveraging their inherent load-following capabilities. This study examines the characteristics of an HPR with closed Brayton Cycle (CBC), and develops a load-following control algorithm. A simplified CBC model is proposed to facilitate control strategy analysis. Model predictive control (MPC) is employed to suppress temperature fluctuations, revealing that the dynamic response of output power under MPC resembles that of a first-order inertial system. Consequently, a power control algorithm based on first-order inertial feedforward control is introduced. Simulation results demonstrate that the proposed algorithm, with a time constant ranging between 500 and 1000 s, significantly mitigates temperature and power fluctuations in HPRs during load-following dynamic power regulation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110986"},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526945","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}
This study presents the results of testing nuclear data libraries by analyzing statistical criteria obtained from comparing experimental and calculated rates for (n,2n), (n,p), (n,pn), (n,nꞌγ) (n,α) and (n,γ) reactions measured on samples natNi, natZr, natNb, natCd, natTi, natCo,63(96%), 65(99.70%)Cu, 64(99.70%)Zn, natIn, natAl, natMg, natFe, natAu and natTh, which were placed in the experimental channels of micromodels of the fusion blanket.
The “fast” (the cylinder Ø 230 mm and 520 mm length was filled with ∼ 67 kg of molten salt 0.52NaF + 0.48ZrF4) and the “thermal” blanket (the same cylinder was placed in a dry channel inside a cubic container filled with water with dimensions of 52.0 × 52.0 × 52.0 cm were investigated. The reaction rates were measured using the activation method.
Modeling with transport codes MCNP5, KIR, PHITS-3.31, SuperMC3.4.0 was performed using the ENDF/B-VII.0 library for neutron transport as well as seven neutron data libraries for reaction rates simulation, including: JEFF-3.3, JENDL-4.0, ENDF/B–VIII.0, ROSFOND-2010, FENDL-3.0, TENDL − 2019 and IRDFF-II.
{"title":"Verification of nuclear data libraries used to design molten salt blankets of a fusion neutron source","authors":"Yu.E. Titarenko, S.A. Balyuk, V.F. Batyaev, V.I. Belousov, I.A. Bedretdinov, V. Yu. Blandinskiy, V.D. Davidenko, I.I. Dyachkov, V.M. Zhivun, Ya.O. Zaritstkiy, M.V. Ioannisian, A.S. Kirsanov, A.A. Kovalishin, N.A. Kovalenko, B.V. Kuteev, V.O. Legostaev, M.R. Malkov, I.V. Mednikov, K.V. Pavlov, A. Yu. Titarenko, K.G. Chernov","doi":"10.1016/j.anucene.2024.110983","DOIUrl":"10.1016/j.anucene.2024.110983","url":null,"abstract":"<div><div>This study presents the results of testing nuclear data libraries by analyzing statistical criteria obtained from comparing experimental and calculated rates for (n,2n), (n,p), (n,pn), (n,nꞌγ) (n,α) and (n,γ) reactions measured on samples <sup>nat</sup>Ni, <sup>nat</sup>Zr, <sup>nat</sup>Nb, <sup>nat</sup>Cd, <sup>nat</sup>Ti, <sup>nat</sup>Co,<sup>63(96%), 65(99.70%)</sup>Cu, <sup>64(99.70%)</sup>Zn, <sup>nat</sup>In, <sup>nat</sup>Al, <sup>nat</sup>Mg, <sup>nat</sup>Fe, <sup>nat</sup>Au and <sup>nat</sup>Th, which were placed in the experimental channels of micromodels of the fusion blanket.</div><div>The “fast” (the cylinder Ø 230 mm and 520 mm length was filled with ∼ 67 kg of molten salt 0.52NaF + 0.48ZrF4) and the “thermal” blanket (the same cylinder was placed in a dry channel inside a cubic container filled with water with dimensions of 52.0 × 52.0 × 52.0 cm were investigated. The reaction rates were measured using the activation method.</div><div>Modeling with transport codes MCNP5, KIR, PHITS-3.31, SuperMC3.4.0 was performed using the ENDF/B-VII.0 library for neutron transport as well as seven neutron data libraries for reaction rates simulation, including: JEFF-3.3, JENDL-4.0, ENDF/B–VIII.0, ROSFOND-2010, FENDL-3.0, TENDL − 2019 and IRDFF-II.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110983"},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527012","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 : 2024-10-21DOI: 10.1016/j.anucene.2024.110997
Emil Fridman , Jacob D. Smith , Dan Kotlyar
This study explores the calculation of Reference Discontinuity Factors (RDFs) using the Serpent Monte Carlo code, focusing on the methodology and potential pitfalls. In two-step reactor analyses, consistently generated RDFs are crucial for aligning homogeneous nodal diffusion results with the reference heterogeneous transport solution. However, the Serpent internal diffusion solver, based on the Analytic Function Expansion Nodal (AFEN) method, may not be compatible with other nodal methods such as the Nodal Expansion Method (NEM). Additionally, the solver can suffer from instabilities, particularly in multi-group calculations, leading to erroneous RDFs. Despite these challenges, Serpent can generate the necessary raw data for RDF calculation, which can be accurately processed using external diffusion solvers. Two numerical examples − a 1D fuel-reflector model and a 2D SMR core model − illustrate the effects of consistent and inconsistent RDFs on simulation accuracy. The study emphasizes the importance of using compatible diffusion solvers and thoroughly assessing RDFs to avoid errors in reactor simulations.
本研究探讨了使用 Serpent Monte Carlo 代码计算参考不连续因子 (RDF)，重点是计算方法和潜在误区。在两步反应器分析中，一致生成的 RDF 对于使均质节点扩散结果与参考异质输运解决方案保持一致至关重要。然而，基于解析函数展开节点法（AFEN）的蛇形内部扩散求解器可能与节点展开法（NEM）等其他节点法不兼容。此外，该求解器可能会出现不稳定的情况，特别是在多组计算中，从而导致错误的 RDF。尽管存在这些挑战，Serpent 仍能生成 RDF 计算所需的原始数据，并使用外部扩散求解器对其进行精确处理。两个数值实例--1D 燃料反射器模型和 2D SMR 核心模型--说明了一致和不一致的 RDF 对模拟精度的影响。该研究强调了使用兼容的扩散求解器和彻底评估 RDF 以避免反应堆模拟错误的重要性。
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