The author independently conducted a detailed analysis of plant conditions, based on a new meltdown process, during the 14 hours after all power loss at the Fukushima Daiichi Nuclear Power Plant Unit 1. After 3 ： 30 PM on March 11, 2011, all power loss caused reactor core cooling function loss and the vaporization of water in the reactor vessel （ RPV ） , and the RPV was filled with high-temperature hydrogen gas generated by decay heat and Zr - H 2 O reactions. The reactor core melted and the periph-eral stainless-steel structures and nuclear vessel walls were heated by the radiation energy of decay heat. Then, the shrouds and other components near the reactor core started melting and the wall temperature of the RPV was raised. Along with the wall temperature rising to nearly 650 ℃ , the aluminum insulation near the wall started melting （ aluminum melting point, about 650 ℃） . The heat from the collapsed insulation caused the superheated state of the containment pressure and temperature of 0.84 MPa and 400 ℃ , respectively, around 3 AM from the saturated state of 0.6 MPa around 1 AM on March 12. After that, the containment vessel was depressurized gradually and kept under stable cooling 14 hours after all power loss, for the time from 4 to 6 AM. By the first water injection to the reac tor core at 4 AM, radioactivity release increased slightly and the containment pressure was stable. However, the situation changed significantly after continuous water injection to the reactor core start ed at around 6 AM. The water injection into the high-temperature reactor core generated film boiling （ poor heat transfer ） and good reactor core cooling was no longer available; thus, increased heating by the Zr - H 2 O reactions started. The results of the author’s analysis showed good consistency with the measured reactor pressure, containment pressure, containment temperature and radioactivity near the main gate. Although decay heat cannot be decreased intentionally, the occurrence of the Zr - H 2 O reactions can be inhibited. It is proposed that the best way to mitigate the effects of a meltdown is to stop water injection to the core after detecting the initiation of film boiling or inferring that the fuel rods are not covered with water.
{"title":"Analysis of Status Fourteen Hours after All Power Loss at the Fukushima Daiichi Nuclear Power Plant Unit 1","authors":"Tsuyoshi Matsuoka","doi":"10.3327/taesj.j21.006","DOIUrl":"https://doi.org/10.3327/taesj.j21.006","url":null,"abstract":"The author independently conducted a detailed analysis of plant conditions, based on a new meltdown process, during the 14 hours after all power loss at the Fukushima Daiichi Nuclear Power Plant Unit 1. After 3 ： 30 PM on March 11, 2011, all power loss caused reactor core cooling function loss and the vaporization of water in the reactor vessel （ RPV ） , and the RPV was filled with high-temperature hydrogen gas generated by decay heat and Zr - H 2 O reactions. The reactor core melted and the periph-eral stainless-steel structures and nuclear vessel walls were heated by the radiation energy of decay heat. Then, the shrouds and other components near the reactor core started melting and the wall temperature of the RPV was raised. Along with the wall temperature rising to nearly 650 ℃ , the aluminum insulation near the wall started melting （ aluminum melting point, about 650 ℃） . The heat from the collapsed insulation caused the superheated state of the containment pressure and temperature of 0.84 MPa and 400 ℃ , respectively, around 3 AM from the saturated state of 0.6 MPa around 1 AM on March 12. After that, the containment vessel was depressurized gradually and kept under stable cooling 14 hours after all power loss, for the time from 4 to 6 AM. By the first water injection to the reac tor core at 4 AM, radioactivity release increased slightly and the containment pressure was stable. However, the situation changed significantly after continuous water injection to the reactor core start ed at around 6 AM. The water injection into the high-temperature reactor core generated film boiling （ poor heat transfer ） and good reactor core cooling was no longer available; thus, increased heating by the Zr - H 2 O reactions started. The results of the author’s analysis showed good consistency with the measured reactor pressure, containment pressure, containment temperature and radioactivity near the main gate. Although decay heat cannot be decreased intentionally, the occurrence of the Zr - H 2 O reactions can be inhibited. It is proposed that the best way to mitigate the effects of a meltdown is to stop water injection to the core after detecting the initiation of film boiling or inferring that the fuel rods are not covered with water.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437728","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}
To analyze the radioactive wastes produced by the accident at the Fukushima Daiichi Nuclear Power Plant, we have been developing a laser absorption and emission spectroscopic system for sample screening. By the sputtering technique, neutral atomic vapor can be produced from a solid sample even if the chemical forms of the samples are complex. This system utilizes a DC glow-discharged hollow cathode atomic source that does not require an extensive chemical pretreatment. Although it is possible to perform emission spectroscopy for multi-element simultaneous analysis to clarify the ele-ments of samples with this system, it takes much time to assign complex peaks of the samples. A program was developed for the rapid assignment of the emission spectra. We utilized a Czerny – Turner spectrometer, which has a low resolution power. The program written in Python could assign the large emission peaks because of the ease of signal discrimination.
{"title":"Development of Element Assignment Program Based on Feature Quantity Extraction Method from Emission Spectra of Glow-Discharged Atomic Vapor Source Obtained with Czerny–Turner Spectrometer","authors":"D. Ishikawa, N. Oku, S. Hasegawa","doi":"10.3327/taesj.j20.012","DOIUrl":"https://doi.org/10.3327/taesj.j20.012","url":null,"abstract":"To analyze the radioactive wastes produced by the accident at the Fukushima Daiichi Nuclear Power Plant, we have been developing a laser absorption and emission spectroscopic system for sample screening. By the sputtering technique, neutral atomic vapor can be produced from a solid sample even if the chemical forms of the samples are complex. This system utilizes a DC glow-discharged hollow cathode atomic source that does not require an extensive chemical pretreatment. Although it is possible to perform emission spectroscopy for multi-element simultaneous analysis to clarify the ele-ments of samples with this system, it takes much time to assign complex peaks of the samples. A program was developed for the rapid assignment of the emission spectra. We utilized a Czerny – Turner spectrometer, which has a low resolution power. The program written in Python could assign the large emission peaks because of the ease of signal discrimination.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437552","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}
Before the Fukushima-Daiichi Accident, Japanese nuclear power plants （ NPPs ） were utilized with a capacity factor of around 70 ％ , 20 ％ lower than the US capacity factor of around 90 ％ , which is a consequence of Japanese NPPs being operated with a shorter fuel cycle and longer outage period. One reason for this situation is that Japanese decision making is strongly focused on equipment reli-ability. In a typical pressurized-water reactor, however, core damage frequency （ CDF ） during refueling outage is higher than that during operation, that is, a short fuel cycle could possibly increase the total CDF of NPPs. In this paper, a decision-making rule using an index representing CDF par power generation is firstly proposed. Secondly, using this rule, a decision process is simulated to optimize the fuel cycle and refueling outage period while showing the effects on CDF reduction in each plant. Thirdly, by applying this decision process to all Japanese NPPs, the total CDF reduction in Japan is indicated. This simulation shows that the change of decision-making rule will bring about an 18 ％ CDF reduction or 16 ％ increase in power generation in total in Japan. At the same time, each NPP gains strong incentive to improve its own safety because this new rule permits a higher capacity factor operation only for the NPPs that are safer than the average.
{"title":"Practical Method of Risk-Informed Decision Making in Nuclear Power Plant Operation and Its Safety Effect","authors":"Yutaka Shikami","doi":"10.3327/TAESJ.J19.029","DOIUrl":"https://doi.org/10.3327/TAESJ.J19.029","url":null,"abstract":"Before the Fukushima-Daiichi Accident, Japanese nuclear power plants （ NPPs ） were utilized with a capacity factor of around 70 ％ , 20 ％ lower than the US capacity factor of around 90 ％ , which is a consequence of Japanese NPPs being operated with a shorter fuel cycle and longer outage period. One reason for this situation is that Japanese decision making is strongly focused on equipment reli-ability. In a typical pressurized-water reactor, however, core damage frequency （ CDF ） during refueling outage is higher than that during operation, that is, a short fuel cycle could possibly increase the total CDF of NPPs. In this paper, a decision-making rule using an index representing CDF par power generation is firstly proposed. Secondly, using this rule, a decision process is simulated to optimize the fuel cycle and refueling outage period while showing the effects on CDF reduction in each plant. Thirdly, by applying this decision process to all Japanese NPPs, the total CDF reduction in Japan is indicated. This simulation shows that the change of decision-making rule will bring about an 18 ％ CDF reduction or 16 ％ increase in power generation in total in Japan. At the same time, each NPP gains strong incentive to improve its own safety because this new rule permits a higher capacity factor operation only for the NPPs that are safer than the average.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437408","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}
2021 ） A new method of evaluating colloid diffusion and filtration in compacted bentonites using den drimers was developed. Polyamidoamine （ PAMAM ） dendrimers with sizes of 5.7 and 7.2 nm were chosen as diffusion probes owing to their high monodispersity and well-defined molecular structure. The diffusion and filtration behaviors were investigated by a through-diffusion experiment in benton ite compacted to 0.8 Mg / m 3 and saturated with 0.005–0.5 mol / L NaCl. The breakthrough curves were observed under all conditions, demonstrating that dendrimers could diffuse through the pore network in compacted bentonite. The effective diffusivity （ D e ） and filtration ratio （ R f ） of dendrimers were de-termined from the breakthrough curves and the depth profiles in compacted bentonite, respectively. The D e values of the negatively charged dendrimers increased when porewater salinity increased and dendrimer size decreased owing to the anion exclusion effect on negatively charged clay surfaces. The R f values increased when porewater salinity decreased and dendrimer size increased, demonstrating that significant fractions of dendrimers were filtered by the narrow pores in the complex pore net -works. The results of this study confirm the validity of the evaluation method using dendrimers and the importance of the need for further investigation under various conditions to understand the fac tors controlling colloid diffusion and filtration and their relationships with the microstructure in com pacted bentonite.
2021)开发了一种利用孔洞驱动剂评价膨润土中胶体扩散和过滤的新方法。选用尺寸为5.7 nm和7.2 nm的聚胺胺(PAMAM)树状大分子作为扩散探针，因为其具有较高的单分散性和清晰的分子结构。通过透扩散实验，研究了膨润土在压实至0.8 Mg / m 3、0.005 ~ 0.5 mol / L NaCl饱和条件下的扩散和过滤行为。在所有条件下观察到的突破曲线表明，树状大分子可以在压实的膨润土中通过孔隙网络扩散。树状大分子的有效扩散系数(D e)和过滤比(R f)分别由压实膨润土的穿透曲线和深度剖面确定。带负电荷的树状大分子的D e值随着孔隙水矿化度的增加而增加，并且由于带负电荷的粘土表面的阴离子排斥作用，树状大分子的尺寸减小。随着孔隙水矿化度的降低和枝状大分子尺寸的增大，R - f值增大，表明在复杂的孔隙网络中，相当一部分枝状大分子被狭窄的孔隙过滤了。本研究的结果证实了树状大分子评价方法的有效性，以及在不同条件下进一步研究控制胶体扩散和过滤的因素及其与膨润土微观结构的关系的重要性。
{"title":"Development of evaluation method for diffusion and filtration behavior of colloid in compacted bentonites using dendrimers","authors":"T. Endo, Y. Tachi, T. Ishidera, M. Terashima","doi":"10.3327/TAESJ.J20.001","DOIUrl":"https://doi.org/10.3327/TAESJ.J20.001","url":null,"abstract":"2021 ） A new method of evaluating colloid diffusion and filtration in compacted bentonites using den drimers was developed. Polyamidoamine （ PAMAM ） dendrimers with sizes of 5.7 and 7.2 nm were chosen as diffusion probes owing to their high monodispersity and well-defined molecular structure. The diffusion and filtration behaviors were investigated by a through-diffusion experiment in benton ite compacted to 0.8 Mg / m 3 and saturated with 0.005–0.5 mol / L NaCl. The breakthrough curves were observed under all conditions, demonstrating that dendrimers could diffuse through the pore network in compacted bentonite. The effective diffusivity （ D e ） and filtration ratio （ R f ） of dendrimers were de-termined from the breakthrough curves and the depth profiles in compacted bentonite, respectively. The D e values of the negatively charged dendrimers increased when porewater salinity increased and dendrimer size decreased owing to the anion exclusion effect on negatively charged clay surfaces. The R f values increased when porewater salinity decreased and dendrimer size increased, demonstrating that significant fractions of dendrimers were filtered by the narrow pores in the complex pore net -works. The results of this study confirm the validity of the evaluation method using dendrimers and the importance of the need for further investigation under various conditions to understand the fac tors controlling colloid diffusion and filtration and their relationships with the microstructure in com pacted bentonite.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437475","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 purpose of this comment is to clarify the whole history of reactor and containment pressure change during the Fukushima meltdown accident. It is based on a new approach for film boiling, which is sustained after the Zr – H 2 O reaction. As the reaction rate is proportional to the reactor or containment vessel pressure under film boiling, it increases rapidly and stops abruptly while sustaining film boiling. The containment vessel pressure change consists of three phases, namely, pressurization, holding a high pressure and depressurization. The containment vessel is pressurized with H 2 gas and steam produced by the Zr – H 2 O reaction and depressurized by heat removal by heatsinks such as the containment vessel wall and inner concrete after the reaction stops. The high pressure between these pressure changes is sustained by balancing the amount of H 2 gas produced by the reaction and that of gas leaking from the gap of the top hat of the containment vessel. The amount of core decay heat is large, but the change of this is negligible. Thus, pressurization is calculated from the amounts of H 2 gas and steam produced by the Zr – H 2 O reaction. The amount removed by the heatsink balances with that produced by the reaction during the high-pressure phase. Depressurization occurs after the reac tion is over, so the reaction heat rate can be calculated from the heat removal rate of the heatsink, which is equal to the condensation rate during depressurization. The rate of gas leakage can be calcu lated from the reaction rate. It is very important that the reaction rate was slow owing to the insuffi cient steam supply, as the melted core in the Fukushima accident was covered with H 2 gas and steam at a pressure of 0.8 MPa or lower. This is different from the rate （ at approximately 7 MPa ） in the Three Mile Island accident, as the specific volume of steam at 0.8 MPa is ten times larger than that at 7 MPa. The calculation results based on this assumption show that almost all the Zr in each core of Units 1, 2 and 3 reacted with water. The location of a small penetration hole produced by the contact of the high-temperature H 2 gas with the suppression chamber wall, is estimated in Unit 2.
{"title":"New approach for describing reactor and containment pressure change after loss of core cooling at Fukushima meltdown accident","authors":"Tsuyoshi Matsuoka","doi":"10.3327/taesj.j20.033","DOIUrl":"https://doi.org/10.3327/taesj.j20.033","url":null,"abstract":"The purpose of this comment is to clarify the whole history of reactor and containment pressure change during the Fukushima meltdown accident. It is based on a new approach for film boiling, which is sustained after the Zr – H 2 O reaction. As the reaction rate is proportional to the reactor or containment vessel pressure under film boiling, it increases rapidly and stops abruptly while sustaining film boiling. The containment vessel pressure change consists of three phases, namely, pressurization, holding a high pressure and depressurization. The containment vessel is pressurized with H 2 gas and steam produced by the Zr – H 2 O reaction and depressurized by heat removal by heatsinks such as the containment vessel wall and inner concrete after the reaction stops. The high pressure between these pressure changes is sustained by balancing the amount of H 2 gas produced by the reaction and that of gas leaking from the gap of the top hat of the containment vessel. The amount of core decay heat is large, but the change of this is negligible. Thus, pressurization is calculated from the amounts of H 2 gas and steam produced by the Zr – H 2 O reaction. The amount removed by the heatsink balances with that produced by the reaction during the high-pressure phase. Depressurization occurs after the reac tion is over, so the reaction heat rate can be calculated from the heat removal rate of the heatsink, which is equal to the condensation rate during depressurization. The rate of gas leakage can be calcu lated from the reaction rate. It is very important that the reaction rate was slow owing to the insuffi cient steam supply, as the melted core in the Fukushima accident was covered with H 2 gas and steam at a pressure of 0.8 MPa or lower. This is different from the rate （ at approximately 7 MPa ） in the Three Mile Island accident, as the specific volume of steam at 0.8 MPa is ten times larger than that at 7 MPa. The calculation results based on this assumption show that almost all the Zr in each core of Units 1, 2 and 3 reacted with water. The location of a small penetration hole produced by the contact of the high-temperature H 2 gas with the suppression chamber wall, is estimated in Unit 2.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437648","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}
Some of the Cs inside the Fukushima Daiichi Nuclear Power Station would be deposited in chemical forms such as CsI and Cs 2 MoO 4 . Since Cs compounds are generally water-soluble, it is pre-dicted that the migration of Cs through the aqueous phase occurs in the long term. Knowledge of the solubility in water is required as basic data for such migration behavior evaluation. Therefore, this study was conducted to investigate the dissolution properties of CsI and Cs 2 MoO 4 in water at 20 ℉ and 25 ℉ . The solubilities of CsI at 25 ℉ calculated using thermodynamic data and the Pitzer ion interaction model were in good agreement with the literature value. It was found that the literature value of CsI at around room temperature is highly reliable. The experimental value of CsI at 20 ℃ obtained by the OECD test guideline 105 flask method （ test guideline ） was also in good agreement with the literature value. The measured solubility of Cs 2 MoO 4 was 256.8 ± 12.7 （ g / 100 g H 2 O ） at 20 ℃ using the test guideline. This measured solubility of Cs 2 MoO 4 was found to be comparable to those of other alkaline molybdates and considered to be more reliable than the literature value.
福岛第一核电站(Fukushima Daiichi Nuclear Power Station)内部的一些铯将以CsI和c2mo_4等化学形式沉积。由于Cs化合物通常是水溶性的，因此可以预测Cs在水相中的迁移是长期发生的。了解在水中的溶解度是评价这种迁移行为的基础数据。因此，本文研究了CsI和c2moo4在20℉和25℉水中的溶解特性。利用热力学数据和Pitzer离子相互作用模型计算的25℉时CsI的溶解度与文献值吻合较好。研究发现，常温前后CsI的文献值具有较高的可靠性。采用OECD试验指南105烧瓶法(试验指南)得到的20℃下CsI的实验值也与文献值吻合较好。在20℃条件下，c2moo4的溶解度为256.8±12.7 (g / 100 g h2o)。c2moo4的溶解度与其他碱性钼酸盐的溶解度相当，比文献值更可靠。
{"title":"Study on solubility of cesium iodide and cesium molybdate in water at around room temperature","authors":"Jumpei Imoto, K. Nakajima, M. Osaka","doi":"10.3327/taesj.j20.037","DOIUrl":"https://doi.org/10.3327/taesj.j20.037","url":null,"abstract":"Some of the Cs inside the Fukushima Daiichi Nuclear Power Station would be deposited in chemical forms such as CsI and Cs 2 MoO 4 . Since Cs compounds are generally water-soluble, it is pre-dicted that the migration of Cs through the aqueous phase occurs in the long term. Knowledge of the solubility in water is required as basic data for such migration behavior evaluation. Therefore, this study was conducted to investigate the dissolution properties of CsI and Cs 2 MoO 4 in water at 20 ℉ and 25 ℉ . The solubilities of CsI at 25 ℉ calculated using thermodynamic data and the Pitzer ion interaction model were in good agreement with the literature value. It was found that the literature value of CsI at around room temperature is highly reliable. The experimental value of CsI at 20 ℃ obtained by the OECD test guideline 105 flask method （ test guideline ） was also in good agreement with the literature value. The measured solubility of Cs 2 MoO 4 was 256.8 ± 12.7 （ g / 100 g H 2 O ） at 20 ℃ using the test guideline. This measured solubility of Cs 2 MoO 4 was found to be comparable to those of other alkaline molybdates and considered to be more reliable than the literature value.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437662","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}
increase. the significant risk issues. First the gap between the significance used in traditional Japanese CAP and risk-informed significance is clarified in this paper. Second, a CAP process model with two conditions of potential safety issues depending on whether the cause is removed or not is prepared. This model can explain how the CAP works to suppress the recurrence of safety issues. Third, using this model, the trend that the CDF is increased by the safety issues is parametrically simulated. Simulation results show that the CAP using risk-informed significance can reduce the CDF increase by 40 ％ over 20 years. For the NPPs that have long been using an inappropriate significance scale, the application of risk-informed significance in CAP is one of the most effective actions for sup pressing future CDF.
{"title":"Method of Applying Graded Approach in Corrective Action Program and Its Safety Effects","authors":"Yutaka Shikami","doi":"10.3327/TAESJ.J19.025","DOIUrl":"https://doi.org/10.3327/TAESJ.J19.025","url":null,"abstract":"increase. the significant risk issues. First the gap between the significance used in traditional Japanese CAP and risk-informed significance is clarified in this paper. Second, a CAP process model with two conditions of potential safety issues depending on whether the cause is removed or not is prepared. This model can explain how the CAP works to suppress the recurrence of safety issues. Third, using this model, the trend that the CDF is increased by the safety issues is parametrically simulated. Simulation results show that the CAP using risk-informed significance can reduce the CDF increase by 40 ％ over 20 years. For the NPPs that have long been using an inappropriate significance scale, the application of risk-informed significance in CAP is one of the most effective actions for sup pressing future CDF.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437362","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}
Not just whether nuclear power stations should be restarted, but how local agreement over the restart should be achieved has been controversial issues in Japan since the Fukushima Nuclear Di-saster. In this paper, public attitudes towards local agreement in the case of Hamaoka Nuclear Power Plants are explored with a postal questionnaire survey in Shizuoka prefecture. Through descriptive statistics and factor analysis, the study shows that extending the “local” scale and judgement by ordi-nary citizens is given more support than the conventional local agreement process. Factor analysis reveals three factors behind respondents’ attitudes towards local agreement: “conventional decision-makers”, “narrow localism” and “national interests”. The analysis of the factor scores reveals that attitudes towards local agreement differ depending on attitudes towards the restart of the plant and the prefectural referendum, as well as generation, while no significant difference is found among gen ders and residential areas excluding the second factor. By clarifying the public attitudes towards how local agreement should be made, this study makes a significant step toward the design of a socially more agreeable local agreement.
{"title":"Public Attitudes towards Local Agreement over the Restart of Nuclear Power Plants: A Questionnaire Survey on the Case of Hamaoka","authors":"Tomoyuki Tatsumi, T. Nakazawa","doi":"10.3327/taesj.j20.006","DOIUrl":"https://doi.org/10.3327/taesj.j20.006","url":null,"abstract":"Not just whether nuclear power stations should be restarted, but how local agreement over the restart should be achieved has been controversial issues in Japan since the Fukushima Nuclear Di-saster. In this paper, public attitudes towards local agreement in the case of Hamaoka Nuclear Power Plants are explored with a postal questionnaire survey in Shizuoka prefecture. Through descriptive statistics and factor analysis, the study shows that extending the “local” scale and judgement by ordi-nary citizens is given more support than the conventional local agreement process. Factor analysis reveals three factors behind respondents’ attitudes towards local agreement: “conventional decision-makers”, “narrow localism” and “national interests”. The analysis of the factor scores reveals that attitudes towards local agreement differ depending on attitudes towards the restart of the plant and the prefectural referendum, as well as generation, while no significant difference is found among gen ders and residential areas excluding the second factor. By clarifying the public attitudes towards how local agreement should be made, this study makes a significant step toward the design of a socially more agreeable local agreement.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437489","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}
It is necessary to consider the effect of photoneutrons produced by photonuclear reactions in the shielding calculation of a medical linac room with incident electron energies greater than 10 MeV. For photonuclear reaction files, the validation of photoneutron energy distributions has not been reported. We compared the evaluated data of photoneutron energy distributions in LA150 and JENDL / PD-2016.1 with the experimental data for 11 nuclides taken from the EXFOR database. For dominant shielding materials of linac, we validated the tendency of energy distributions by comparing the experimental data of neighborhood nuclides. Consequently, we found problems in the evaluations of the photoneutron energy distributions of the photonuclear reaction files.
{"title":"Validation of Photoneutron Energy Distributions in Shielding Calculation of the Medical Linac Room","authors":"K. Kosako","doi":"10.3327/taesj.j20.016","DOIUrl":"https://doi.org/10.3327/taesj.j20.016","url":null,"abstract":"It is necessary to consider the effect of photoneutrons produced by photonuclear reactions in the shielding calculation of a medical linac room with incident electron energies greater than 10 MeV. For photonuclear reaction files, the validation of photoneutron energy distributions has not been reported. We compared the evaluated data of photoneutron energy distributions in LA150 and JENDL / PD-2016.1 with the experimental data for 11 nuclides taken from the EXFOR database. For dominant shielding materials of linac, we validated the tendency of energy distributions by comparing the experimental data of neighborhood nuclides. Consequently, we found problems in the evaluations of the photoneutron energy distributions of the photonuclear reaction files.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437995","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}
Hiroki Shibahara, Mami Yuki, Yuki Ohishi, Yasutaka Takeuchi, Z. Yoshida
The mass discrimination effect in the isotope analyses of barium isotopes with natural abundance, from Ba-130 to Ba-138, was investigated by triple-quadrupole inductively coupled plasmamass spectrometry (ICP-QQQ). The mass bias coefficients for Ba isotopes, denoted by ε(Ba), were determined from the slope of a linear relationship between the atomic mass differences and the ratios of ion count rates for a given isotope pair experimentally determined and calculated from natural abundances. The effect of the addition of a collision-reaction cell (CRC) gas such as helium, hydrogen, nitrous oxide, or carbon dioxide on ε(Ba) was examined. Large ε(Ba) values were observed in the case of Heor H2-CRC gas, and the values were from +0.8 to +1.7% per atomic mass unit. On the other hand, ε(Ba) observed with CRC gasses containing N2O or CO2 was relatively small and below +0.3% per atomic mass unit. In addition, the dependence of the energy discrimination potential (ED) applied between the CRC and the second quadrupole mass separator of ICP-QQQ on ε(Ba) was investigated. Finally, the analytical mass bias coefficient of the radioactive cesium nuclides Cs, Cs, and Cs, ε(Cs), was discussed for ε(Ba) in the same mass range as that of Ba isotopes.
{"title":"Mass Discrimination Effect in Isotope Ratio Measurement of Barium by Triple-Quadrupole Inductively Coupled Plasma-Mass Spectrometry: Estimation of Mass Bias Coefficient for Isotopic Analysis of Radioactive Cesium Nuclides","authors":"Hiroki Shibahara, Mami Yuki, Yuki Ohishi, Yasutaka Takeuchi, Z. Yoshida","doi":"10.3327/taesj.j19.030","DOIUrl":"https://doi.org/10.3327/taesj.j19.030","url":null,"abstract":"The mass discrimination effect in the isotope analyses of barium isotopes with natural abundance, from Ba-130 to Ba-138, was investigated by triple-quadrupole inductively coupled plasmamass spectrometry (ICP-QQQ). The mass bias coefficients for Ba isotopes, denoted by ε(Ba), were determined from the slope of a linear relationship between the atomic mass differences and the ratios of ion count rates for a given isotope pair experimentally determined and calculated from natural abundances. The effect of the addition of a collision-reaction cell (CRC) gas such as helium, hydrogen, nitrous oxide, or carbon dioxide on ε(Ba) was examined. Large ε(Ba) values were observed in the case of Heor H2-CRC gas, and the values were from +0.8 to +1.7% per atomic mass unit. On the other hand, ε(Ba) observed with CRC gasses containing N2O or CO2 was relatively small and below +0.3% per atomic mass unit. In addition, the dependence of the energy discrimination potential (ED) applied between the CRC and the second quadrupole mass separator of ICP-QQQ on ε(Ba) was investigated. Finally, the analytical mass bias coefficient of the radioactive cesium nuclides Cs, Cs, and Cs, ε(Cs), was discussed for ε(Ba) in the same mass range as that of Ba isotopes.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437464","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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