Raj Kumar Singh, Deb Mukhopadhyay, D. Khakhar, J. B. Joshi
Sub-channel analysis codes are presently a requirement for design and safety analysis of nuclear reactors. Among the crucial inputs for these codes, the turbulent mixing factor holds particular significance. However, acquiring this factor through experimental means proves to be a challenging endeavor, primarily due to the necessity for precise pressure equilibrium between sub-channels. Consequently, this requirement leads to the undertaking of expensive and intricate experiments for each new reactor or in cases where there are modifications in fuel bundle design. The need for Direct Numerical Simulation (DNS) stems from the challenges and costs involved in experimental techniques, and the uncertainties due to empiricism in Computational Fluid Dynamics (CFD) models. In this study, DNS has been conducted across six Reynolds numbers, ranging from 17640 to 1.176×105, in the geometry of a Pressurized Water Reactor (PWR) sub-channel. The resulting turbulent flow structures have been computed and their dynamics is examined. Furthermore, this paper presents a methodology for directly calculating the turbulent mixing factor from the fluctuating velocity field obtained from DNS data. The turbulent mixing process has been scrutinized in-depth, and a correlation for the turbulent mixing factor is developed. It is noted that most of the mixing occurs in the near-wall region. The study suggests different mixing factors for mass and momentum mixing. This paper aims to provide a comprehensive insight into the turbulent mixing phenomenon.
子通道分析代码是目前核反应堆设计和安全分析的必要条件。在这些代码的关键输入中，湍流混合因子具有特别重要的意义。然而，通过实验手段获取该因子被证明是一项具有挑战性的工作，这主要是由于子通道之间必须实现精确的压力平衡。因此，这一要求导致每一个新反应堆或燃料束设计发生变化时，都需要进行昂贵而复杂的实验。直接数值模拟（DNS）的需求源于实验技术所涉及的挑战和成本，以及计算流体动力学（CFD）模型中经验主义所带来的不确定性。在这项研究中，在压水堆 (PWR) 子通道的几何形状中，对从 17640 到 1.176×105 的六个雷诺数进行了 DNS 模拟。本文对由此产生的湍流结构进行了计算，并对其动力学特性进行了研究。此外，本文还介绍了一种从 DNS 数据获得的波动速度场直接计算湍流混合因子的方法。本文深入研究了湍流混合过程，并建立了湍流混合因子的相关关系。研究指出，大部分混合发生在近壁区域。研究提出了质量混合和动量混合的不同混合因子。本文旨在提供对湍流混合现象的全面见解。
{"title":"Estimation of Turbulent Mixing Factor and Study of Turbulent Flow Structures in PWR Sub Channel by DNS","authors":"Raj Kumar Singh, Deb Mukhopadhyay, D. Khakhar, J. B. Joshi","doi":"10.1115/1.4066001","DOIUrl":"https://doi.org/10.1115/1.4066001","url":null,"abstract":"\u0000 Sub-channel analysis codes are presently a requirement for design and safety analysis of nuclear reactors. Among the crucial inputs for these codes, the turbulent mixing factor holds particular significance. However, acquiring this factor through experimental means proves to be a challenging endeavor, primarily due to the necessity for precise pressure equilibrium between sub-channels. Consequently, this requirement leads to the undertaking of expensive and intricate experiments for each new reactor or in cases where there are modifications in fuel bundle design. The need for Direct Numerical Simulation (DNS) stems from the challenges and costs involved in experimental techniques, and the uncertainties due to empiricism in Computational Fluid Dynamics (CFD) models.\u0000 In this study, DNS has been conducted across six Reynolds numbers, ranging from 17640 to 1.176×105, in the geometry of a Pressurized Water Reactor (PWR) sub-channel. The resulting turbulent flow structures have been computed and their dynamics is examined.\u0000 Furthermore, this paper presents a methodology for directly calculating the turbulent mixing factor from the fluctuating velocity field obtained from DNS data. The turbulent mixing process has been scrutinized in-depth, and a correlation for the turbulent mixing factor is developed. It is noted that most of the mixing occurs in the near-wall region. The study suggests different mixing factors for mass and momentum mixing. This paper aims to provide a comprehensive insight into the turbulent mixing phenomenon.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2024-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141820575","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The influence of radial neutron reflector on the build-up and propagation of a nuclear fuel burn-up wave in a fast multiplying medium is investigated using a consistent parametric approach. Coupled multi-group neutron diffusion equations with burn-up evolution model are simulated on the two-dimensional cylindrical reactor geometry with azimuthal symmetry. Uranium-Plutonium transmutation model is considered, and the simulation is performed by using finite element multiphysics software package COMSOL. Transient characteristics of the burn-up wave are represented by two new parameters, namely, Transient Time (TT) and Transient Length (TL). TT and TL are defined as the time and distance required for the burn-up wave to attain its steady-state nature. Steady-state phases are characterized in terms of wave velocity, Full Width Half Maximum (FWHM), and 10% of Maximum (FW10M). A sensitivity study of steady-state and transient parameters is conducted for the different values of radial reflector thickness. The potential relevance of these characterization parameters on the development of optimal geometrical configuration of radial neutron reflector in B&B based reactor design is addressed based on the sensitivity study.
{"title":"Effect of Radial Neutron Reflector on the Characteristics of Nuclear Fuel Burn-up Wave in a Fast Neutron Energy Spectrum Multiplying Medium: A Consistent Parametric Approach","authors":"Dipanjan Ray, S. Saraswat, V. S. Bhadouria","doi":"10.1115/1.4065893","DOIUrl":"https://doi.org/10.1115/1.4065893","url":null,"abstract":"\u0000 The influence of radial neutron reflector on the build-up and propagation of a nuclear fuel burn-up wave in a fast multiplying medium is investigated using a consistent parametric approach. Coupled multi-group neutron diffusion equations with burn-up evolution model are simulated on the two-dimensional cylindrical reactor geometry with azimuthal symmetry. Uranium-Plutonium transmutation model is considered, and the simulation is performed by using finite element multiphysics software package COMSOL. Transient characteristics of the burn-up wave are represented by two new parameters, namely, Transient Time (TT) and Transient Length (TL). TT and TL are defined as the time and distance required for the burn-up wave to attain its steady-state nature. Steady-state phases are characterized in terms of wave velocity, Full Width Half Maximum (FWHM), and 10% of Maximum (FW10M). A sensitivity study of steady-state and transient parameters is conducted for the different values of radial reflector thickness. The potential relevance of these characterization parameters on the development of optimal geometrical configuration of radial neutron reflector in B&B based reactor design is addressed based on the sensitivity study.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2024-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141678122","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents the results of neutron detection efficiency and dosimetry between a borated centrifugally tensioned metastable fluid detector (B-CTMFD) vs He-3 Ludlum-42-49B, and, the Fuji NSN3 conducted using a Cf-252 neutron source behind lead and concrete shielding. MCNP code simulations accounted for 3-D effects and derived cpm/mic.Sv/h factors. Ludlum and NSN3 offer fixed sensitivity, but CTMFD offered on-demand sensitivity by varying its Pneg state between 0.3-0.7 MPa. The B-CTMFD demonstrated sensitivity of up to ~22x greater than Ludlum and 5x greater than NSN3, for 0-15 cm Pb shielding, and 0-30 cm concrete shielding; it overcomes the 60% detection penalty inherent in the NB-CTMF-designed only to detect fast-energy neutrons - as described in the companion (Part-1) paper. Unlike the NB-CTMFD, which used 100% DFP (C5H2F10), the B-CTMFD requires the use of an azeotropic mixture of DFP, methanol, and tri-methyl borate (TMB - using natural boron) in 80:4:16 proportion. The B-CTMFD was about 6 times more sensitive than NB-CTMFD under the most heavily shielded condition and taken together, also offered 2-energy bin neutron spectroscopic enablement, together with 22-5x higher absolute efficiency- relative sensitivity compared with the non-spectroscopic Ludlum (He-3) and NSN3 (methane-nitrogen) based detectors. From an intrinsic efficiency standpoint, the B-CTMFD operating at Pneg = 0.7 MPa state, demonstrated even superior ~103x higher intrinsic efficiency over Ludlum and NSN3.
{"title":"Performance of B-CTMFD Detector Vs Ludlum 42-49B, Fuji NSN3 Detectors for Fission Energy Spectrum Neutron Detection with the Source within Lead/concrete Shielded Configurations","authors":"Yusuke Ota, R. Taleyarkhan","doi":"10.1115/1.4065853","DOIUrl":"https://doi.org/10.1115/1.4065853","url":null,"abstract":"\u0000 This paper presents the results of neutron detection efficiency and dosimetry between a borated centrifugally tensioned metastable fluid detector (B-CTMFD) vs He-3 Ludlum-42-49B, and, the Fuji NSN3 conducted using a Cf-252 neutron source behind lead and concrete shielding. MCNP code simulations accounted for 3-D effects and derived cpm/mic.Sv/h factors. Ludlum and NSN3 offer fixed sensitivity, but CTMFD offered on-demand sensitivity by varying its Pneg state between 0.3-0.7 MPa. The B-CTMFD demonstrated sensitivity of up to ~22x greater than Ludlum and 5x greater than NSN3, for 0-15 cm Pb shielding, and 0-30 cm concrete shielding; it overcomes the 60% detection penalty inherent in the NB-CTMF-designed only to detect fast-energy neutrons - as described in the companion (Part-1) paper. Unlike the NB-CTMFD, which used 100% DFP (C5H2F10), the B-CTMFD requires the use of an azeotropic mixture of DFP, methanol, and tri-methyl borate (TMB - using natural boron) in 80:4:16 proportion. The B-CTMFD was about 6 times more sensitive than NB-CTMFD under the most heavily shielded condition and taken together, also offered 2-energy bin neutron spectroscopic enablement, together with 22-5x higher absolute efficiency- relative sensitivity compared with the non-spectroscopic Ludlum (He-3) and NSN3 (methane-nitrogen) based detectors. From an intrinsic efficiency standpoint, the B-CTMFD operating at Pneg = 0.7 MPa state, demonstrated even superior ~103x higher intrinsic efficiency over Ludlum and NSN3.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2024-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141710877","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Zengkui Zhang, Chongzhi Wu, Zhenguo Peng, Mingkai Li
To ensure effective control of the welding process, nuclear power architecture and engineering (AE) companies routinely undertake the review of Procedure Qualification Records (PQRs) and Welding Procedure Specifications (WPSs) for safety-related structures, systems, and components, as qualified by manufacturers and construction contractors. Given the substantial quantity of PQRs and WPSs necessitating evaluation, along with the multitude of variables inherent in these documents, a set of review checklists has been devised. These checklists, developed from the perspective of AE companies' welding engineers, transform welding procedure qualification rules and requirements into intuitive descriptions, assessments of correctness, or numerical comparisons of scale. AE companies' welding engineers employ these review checklists to scrutinize PQRs and WPSs, facilitating a comprehensive, accurate, and efficient review process.
{"title":"Reviewing Welding Procedures - Checklists for Nuclear Power Systems","authors":"Zengkui Zhang, Chongzhi Wu, Zhenguo Peng, Mingkai Li","doi":"10.1115/1.4065851","DOIUrl":"https://doi.org/10.1115/1.4065851","url":null,"abstract":"\u0000 To ensure effective control of the welding process, nuclear power architecture and engineering (AE) companies routinely undertake the review of Procedure Qualification Records (PQRs) and Welding Procedure Specifications (WPSs) for safety-related structures, systems, and components, as qualified by manufacturers and construction contractors. Given the substantial quantity of PQRs and WPSs necessitating evaluation, along with the multitude of variables inherent in these documents, a set of review checklists has been devised. These checklists, developed from the perspective of AE companies' welding engineers, transform welding procedure qualification rules and requirements into intuitive descriptions, assessments of correctness, or numerical comparisons of scale. AE companies' welding engineers employ these review checklists to scrutinize PQRs and WPSs, facilitating a comprehensive, accurate, and efficient review process.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2024-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141697299","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents results of neutron detection efficiency and dosimetry between a non-borated centrifugally tensioned metastable fluid detector (NB-CTMFD) configured to detect fast (>0.1 MeV) vs. He-3 based Ludlum-42-49B and the Fuji NSN3, detectors behind various levels of lead and concrete shieldin-together with MCNP code simulations to account for 3-D effects and to relate the detection sensitivity with the dose rate. The MCNP results for neutron energy spectra were validated vs. experimental measurements using the H*TMFD. While Ludlum and NSN3 operate at a fixed sensitivity, the NB-CTMFD detector offers variable sensitivity by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The NB-CTMFD (configured for fast: > 0.2 MeV neutron detection) offered relative sensitivity enhancements of up to 15x greater than the Ludlum and ~5x greater than the NSN3 detector for low shielding thicknesses. For larger thicknesses, the advantage factor for NB-CTMFD unit reduces with increased fast neutron removal via down-scattering, which benefits the Ludlum and NSN3 detectors. Our companion paper compares performance with a Cf-252 spectrum using a borated CTMFD (covering dual energy bins: thermal-epithermal and fast energy ranges) wherein, it is demonstrated that the advantage factor for the CTMFD (if borated) remains elevated at the 5-10x higher level for low and large shielding thicknesses and soft and hard neutron spectra.
{"title":"Performance of NB-CTMFD detector vs Ludlum 42-49B, and Fuji NSN3 detectors for hard (Am-Be) and soft (Cf-252 fission) energy spectra neutron sources within lead/concrete shielded configurations","authors":"Yusuke Ota, S. Ozerov, R. Taleyarkhan","doi":"10.1115/1.4065852","DOIUrl":"https://doi.org/10.1115/1.4065852","url":null,"abstract":"This paper presents results of neutron detection efficiency and dosimetry between a non-borated centrifugally tensioned metastable fluid detector (NB-CTMFD) configured to detect fast (>0.1 MeV) vs. He-3 based Ludlum-42-49B and the Fuji NSN3, detectors behind various levels of lead and concrete shieldin-together with MCNP code simulations to account for 3-D effects and to relate the detection sensitivity with the dose rate. The MCNP results for neutron energy spectra were validated vs. experimental measurements using the H*TMFD. While Ludlum and NSN3 operate at a fixed sensitivity, the NB-CTMFD detector offers variable sensitivity by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The NB-CTMFD (configured for fast: > 0.2 MeV neutron detection) offered relative sensitivity enhancements of up to 15x greater than the Ludlum and ~5x greater than the NSN3 detector for low shielding thicknesses. For larger thicknesses, the advantage factor for NB-CTMFD unit reduces with increased fast neutron removal via down-scattering, which benefits the Ludlum and NSN3 detectors. Our companion paper compares performance with a Cf-252 spectrum using a borated CTMFD (covering dual energy bins: thermal-epithermal and fast energy ranges) wherein, it is demonstrated that the advantage factor for the CTMFD (if borated) remains elevated at the 5-10x higher level for low and large shielding thicknesses and soft and hard neutron spectra.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.5,"publicationDate":"2024-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141704991","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The authors goal was to provide a practical guideline on nuclear shelter wall thickness estimation. The shelter wall thickness is assessed on the basis of published data on radiation dose calculations, corroborated by field data from nuclear tests, and information or calculations of the dose attenuation in the shelter wall. Dose criterion is selected as the acceptable dose behind the shield. Doses from nuclear tests or simulations are summarized in tables in relation to bomb yield and distance from the explosion point. Neutron dose transmission factors, defined as ratio of the dose behind the shielding to the dose without shielding, and their interdependence with shield thickness for different materials were found in the references. Based on the required dose transmission factor, calculated with the selected dose criterion, the corresponding shelter concrete wall thickness is assessed, taken into account the relation with the dose transmission factors from the literature. For gamma rays the shield thickness is calculated on the basis of analytical functions describing the dependence between dose without shield, dose criterion and shield thickness. Data tables are provided with the assessed concrete wall thickness in relation to bomb yield and distance.
{"title":"Practical Approach to Nuclear Shelter Wall Thickness Estimation","authors":"Rayna Hristova, Stefan Simovsky","doi":"10.1115/1.4065463","DOIUrl":"https://doi.org/10.1115/1.4065463","url":null,"abstract":"\u0000 The authors goal was to provide a practical guideline on nuclear shelter wall thickness estimation. The shelter wall thickness is assessed on the basis of published data on radiation dose calculations, corroborated by field data from nuclear tests, and information or calculations of the dose attenuation in the shelter wall. Dose criterion is selected as the acceptable dose behind the shield. Doses from nuclear tests or simulations are summarized in tables in relation to bomb yield and distance from the explosion point. Neutron dose transmission factors, defined as ratio of the dose behind the shielding to the dose without shielding, and their interdependence with shield thickness for different materials were found in the references. Based on the required dose transmission factor, calculated with the selected dose criterion, the corresponding shelter concrete wall thickness is assessed, taken into account the relation with the dose transmission factors from the literature. For gamma rays the shield thickness is calculated on the basis of analytical functions describing the dependence between dose without shield, dose criterion and shield thickness. Data tables are provided with the assessed concrete wall thickness in relation to bomb yield and distance.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141007322","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This study assessed the response function of a p-type and n-type coaxial high-purity germanium (HPGe) detector via Monte Carlo simulations. MCNPX was employed to model the Coaxial Ge detectors, and for a precise simulation, the dimensions of the dead layer of germanium crystals were added. The dead layer was separated into front and lateral surfaces, and the thickness of each dead layer was modeled. In this work, the simulated detectors have been performed at different energy lines using a radioactive source Eu-152 to study the response function of each with dead layer variations for the front dead layer and study the range of relative deviation of the Monte Carlo simulation outputs from the manufactured declared data. The results proved that the n-type coaxial high-purity germanium (HPGe) detector is more sensitive to the dead layer change than p-type with a thick change by 0.02 mm. This research has significant effects on the efficiencies of the radiation detection systems in the energy range ~ (120-1410) KeV.
本研究通过蒙特卡洛模拟评估了 p 型和 n 型同轴高纯锗（HPGe）探测器的响应函数。采用 MCNPX 对同轴锗探测器进行建模，并添加了锗晶体死层的尺寸以实现精确模拟。死层被分为正面和侧面，每个死层的厚度都被建模。在这项工作中，利用放射源 Eu-152 在不同的能量线对模拟探测器进行了测试，以研究每个探测器的响应函数与前死层的死层变化，并研究蒙特卡罗模拟输出与制造申报数据的相对偏差范围。结果证明，n 型同轴高纯锗（HPGe）探测器对死层变化的敏感度高于厚度变化 0.02 毫米的 p 型探测器。这项研究对能量范围 ~ (120-1410) KeV 的辐射探测系统的效率有重大影响。
{"title":"Modeling of N and P-type of Coaxial Ge Detectors Using MCNPX and the Effect of Dead Layer Variation on Its Response Function","authors":"R.A. El-Tayebany, Mohamed Ali, Nawal Mohames, Rania Mohamed","doi":"10.1115/1.4065395","DOIUrl":"https://doi.org/10.1115/1.4065395","url":null,"abstract":"\u0000 This study assessed the response function of a p-type and n-type coaxial high-purity germanium (HPGe) detector via Monte Carlo simulations. MCNPX was employed to model the Coaxial Ge detectors, and for a precise simulation, the dimensions of the dead layer of germanium crystals were added. The dead layer was separated into front and lateral surfaces, and the thickness of each dead layer was modeled. In this work, the simulated detectors have been performed at different energy lines using a radioactive source Eu-152 to study the response function of each with dead layer variations for the front dead layer and study the range of relative deviation of the Monte Carlo simulation outputs from the manufactured declared data. The results proved that the n-type coaxial high-purity germanium (HPGe) detector is more sensitive to the dead layer change than p-type with a thick change by 0.02 mm. This research has significant effects on the efficiencies of the radiation detection systems in the energy range ~ (120-1410) KeV.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140662321","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper presents research work focused on assessments for meeting the challenge of monitoring actinide content in spent nuclear fuel (SNF) via characteristic neutron emissions [from spontaneous fission and (α,n) reactions] with CTMFDs. A challenge problem was posed to examine if a CTMFD could operate reliably over 1 hour for conducting neutron spectroscopy at a 1 m standoff from a 30-y cooled SNF, in a ~1012:1 (gamma:neutron) and a 150 Gy (15 kRad) accumulated dose. The impacts on operability were studied for the effects of gamma radiation on: (i) radiolysis in the CTMFD sensing fluid; (ii) 3 MeV gamma photoneutrons; and, (iii) CTMFD electronics. A Co-60 irradiator was used for dose effects on the CTMFD. A 14 MeV DT accelerator was used with a NaCl target to produce 3-4 MeV photons from activated 37S (via. neutron absorption in 37Cl) expected from SNF at 1-m standoff. Our examinations revealed the absence of any significant impact on CTMFD performance for meeting and exceeding the challenge problem metrics. We validated for no discernible impact of: 3-4 MeV gamma-produced photoneutrons when combined with a fission neutron source and radiolysis in the DFP sensing fluid through a 150 Gy absorbed dose. Past research results at Purdue University have validated survivability above the targeted 150 Gy level. This paper also provides extended evidence for survivability (from radiolysis) at higher gamma doses through 750 Gy with a borated DFP-sensing fluid formulation-based CTMFD.
{"title":"Monitoring for Actinide Neutron Emissions From Spent Nuclear Fuel Under Extreme Gamma Fields Using Centrifugally Tensioned Metastable Fluid Detector Sensor Technology","authors":"C. Harabagiu, R. Taleyarkhan","doi":"10.1115/1.4065280","DOIUrl":"https://doi.org/10.1115/1.4065280","url":null,"abstract":"\u0000 This paper presents research work focused on assessments for meeting the challenge of monitoring actinide content in spent nuclear fuel (SNF) via characteristic neutron emissions [from spontaneous fission and (α,n) reactions] with CTMFDs. A challenge problem was posed to examine if a CTMFD could operate reliably over 1 hour for conducting neutron spectroscopy at a 1 m standoff from a 30-y cooled SNF, in a ~1012:1 (gamma:neutron) and a 150 Gy (15 kRad) accumulated dose. The impacts on operability were studied for the effects of gamma radiation on: (i) radiolysis in the CTMFD sensing fluid; (ii) 3 MeV gamma photoneutrons; and, (iii) CTMFD electronics. A Co-60 irradiator was used for dose effects on the CTMFD. A 14 MeV DT accelerator was used with a NaCl target to produce 3-4 MeV photons from activated 37S (via. neutron absorption in 37Cl) expected from SNF at 1-m standoff. Our examinations revealed the absence of any significant impact on CTMFD performance for meeting and exceeding the challenge problem metrics. We validated for no discernible impact of: 3-4 MeV gamma-produced photoneutrons when combined with a fission neutron source and radiolysis in the DFP sensing fluid through a 150 Gy absorbed dose. Past research results at Purdue University have validated survivability above the targeted 150 Gy level. This paper also provides extended evidence for survivability (from radiolysis) at higher gamma doses through 750 Gy with a borated DFP-sensing fluid formulation-based CTMFD.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-04-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140734597","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Accident analysis and ensuring power plant safety are pivotal in the nuclear energy sector. Significant strides have been achieved over the past few decades regarding fire protection and safety, primarily centered on design and regulatory compliance. Yet, after the Fukushima accident a decade ago, the imperative to enhance measures against fire, internal flooding, and power loss has intensified. Hence, a comprehensive, multilayered protection strategy against severe accidents is needed. Consequently, gaining a deeper insight into pool fires and their behavior through extensive validated data can greatly aid in improving these measures using advanced validation techniques. A model validation study was performed at Sandia National Laboratories in which a 30-cm diameter methanol pool fire was modeled using the SIERRA/Fuego turbulent reacting flow code. This validation study used a standard validation experiment to compare model results against, and conclusions have been published. The fire was modeled with a Large Eddy Simulation (LES) turbulence model with subgrid turbulent kinetic energy closure. Combustion was modeled using a strained laminar flamelet library approach. Radiative heat transfer was accounted for with a model utilizing the gray-gas approximation. In the present study, additional validation analysis is performed using the area validation metric (AVM). These activities are done on multiple datasets involving different variables and temporal/spatial ranges and intervals. The results provide insight into the use of the area validation metric on such temporally varying datasets and the importance of physics-aware use of the metric for proper analysis.
{"title":"Validation Assessment of Turbulent Reacting Flow Model Using The Area-Validation Metric on Medium-Scale Methanol Pool Fire Results","authors":"Jared Kirsch, Nima Fathi","doi":"10.1115/1.4065173","DOIUrl":"https://doi.org/10.1115/1.4065173","url":null,"abstract":"\u0000 Accident analysis and ensuring power plant safety are pivotal in the nuclear energy sector. Significant strides have been achieved over the past few decades regarding fire protection and safety, primarily centered on design and regulatory compliance. Yet, after the Fukushima accident a decade ago, the imperative to enhance measures against fire, internal flooding, and power loss has intensified. Hence, a comprehensive, multilayered protection strategy against severe accidents is needed. Consequently, gaining a deeper insight into pool fires and their behavior through extensive validated data can greatly aid in improving these measures using advanced validation techniques. A model validation study was performed at Sandia National Laboratories in which a 30-cm diameter methanol pool fire was modeled using the SIERRA/Fuego turbulent reacting flow code. This validation study used a standard validation experiment to compare model results against, and conclusions have been published. The fire was modeled with a Large Eddy Simulation (LES) turbulence model with subgrid turbulent kinetic energy closure. Combustion was modeled using a strained laminar flamelet library approach. Radiative heat transfer was accounted for with a model utilizing the gray-gas approximation. In the present study, additional validation analysis is performed using the area validation metric (AVM). These activities are done on multiple datasets involving different variables and temporal/spatial ranges and intervals. The results provide insight into the use of the area validation metric on such temporally varying datasets and the importance of physics-aware use of the metric for proper analysis.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140379431","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper proposes to solve the issues and related to Aging Management Program (AMP) during the decommissioning phase by using risk information. The AMP for the decommissioning plants has not been established, while the number of permanently shut down and decommissioned nuclear power plants are increasing in the world. The first issue is the change of the functional importance of the safety systems at the beginning of decommissioning. We propose to evaluate the importance using the PRA model with the change of specification in the decommissioning plant and to determine new importance grade. The second issue is to make the maintenance based on the appropriate risk information. This risk information should be based on the progress of decommissioning, reflected and evaluated in the PRA model, as well as information obtained from maintenance inspections, including the ROP process, and failure experience during routine operation. To address these issues, the maintenance program in decommissioning that uses the evaluation of risk information according to the decommissioning phase is proposed in this paper.
本文提出利用风险信息解决退役阶段老化管理计划（AMP）的相关问题。退役核电站的老化管理计划尚未建立，而世界上永久关闭和退役的核电站数量却在不断增加。第一个问题是退役初期安全系统功能重要性的变化。我们建议使用 PRA 模型评估退役核电厂规格变化后的重要性，并确定新的重要性等级。第二个问题是根据适当的风险信息进行维护。这些风险信息应基于 PRA 模型所反映和评估的退役进度，以及从维护检查（包括 ROP 过程）中获得的信息和日常运行中的故障经验。为解决这些问题，本文提出了根据退役阶段使用风险信息评估的退役维护计划。
{"title":"Development of Risk Informed Aging Management Program in Decommissioning","authors":"Masao Uesaka, Kenta Murakami","doi":"10.1115/1.4065139","DOIUrl":"https://doi.org/10.1115/1.4065139","url":null,"abstract":"\u0000 This paper proposes to solve the issues and related to Aging Management Program (AMP) during the decommissioning phase by using risk information. The AMP for the decommissioning plants has not been established, while the number of permanently shut down and decommissioned nuclear power plants are increasing in the world.\u0000 The first issue is the change of the functional importance of the safety systems at the beginning of decommissioning. We propose to evaluate the importance using the PRA model with the change of specification in the decommissioning plant and to determine new importance grade.\u0000 The second issue is to make the maintenance based on the appropriate risk information. This risk information should be based on the progress of decommissioning, reflected and evaluated in the PRA model, as well as information obtained from maintenance inspections, including the ROP process, and failure experience during routine operation.\u0000 To address these issues, the maintenance program in decommissioning that uses the evaluation of risk information according to the decommissioning phase is proposed in this paper.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2024-03-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140223969","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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