Atomic Energy最新文献

{"title":"Effects of changes in the coolant flow rate of the VVER reactor to the response from the self-powered neutron detectors of the in-core monitoring system: simulation of transient modes in the PRIM-AES software package","authors":"D. A. Gornostaev,&nbsp;A. Yu. Kurchenkov,&nbsp;A. M. Musikhin,&nbsp;Yu. M. Semchenkov","doi":"10.1007/s10512-025-01247-3","DOIUrl":"10.1007/s10512-025-01247-3","url":null,"abstract":"<div><p>In the present paper, we use computational simulation to investigate the response of self-powered neutron detectors (SPNDs) to changes in the coolant flow rate of the measured fuel assembly (FA). A model problem of reducing the coolant flow rate in several measured FAs is considered. Their response is assessed, including the difference in the SPND current before and after introducing a disturbance caused by a decrease in the flow rate through the FA. In some cases, readings of in-core monitoring system SPNDs can be used to diagnose a decrease in the flow rate through the FA.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 4","pages":"212 - 217"},"PeriodicalIF":0.3,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145184097","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Testing of a VVER-1200 reactor in a daily load schedule mode at the 1st power unit of the Leningrad NPP-2","authors":"P. E. Filimonov,&nbsp;Yu. M. Semchenkov,&nbsp;P. P. Mezencev,&nbsp;A. E. Shirvanyants,&nbsp;S. A. Eremeev,&nbsp;A. V. Kalinnikov","doi":"10.1007/s10512-025-01246-4","DOIUrl":"10.1007/s10512-025-01246-4","url":null,"abstract":"<div><p>The paper presents the results of testing the operation of a reactor at the 1st power unit of the Leningrad NPP‑2 in the daily load schedule mode. The tests aimed to experimentally substantiate the operation of the VVER-1200 power unit in a daily load schedule, as well as to develop control methods and check the operation of standard equipment. Two alternative methods for controlling the reactor power density were tested in combination with soft temperature control of reactivity by changing the steam pressure of the secondary circuit. The operation of the boron control system modernized to automatically compensate for reactor xenon poisoning was tested and adjusted. The performed tests confirmed the ability of the VVER-1200 power unit to operate in a daily load schedule both at the beginning and end of the fuel campaign.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 4","pages":"203 - 211"},"PeriodicalIF":0.3,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145184085","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Immobilization of spent sorbent for separation of americium and curium into magnesium potassium phosphate compound","authors":"S. A. Fimina,&nbsp;N. D. Chalysheva,&nbsp;K. Yu. Belova,&nbsp;B. V. Savel’ev,&nbsp;S. E. Vinokurov","doi":"10.1007/s10512-025-01240-w","DOIUrl":"10.1007/s10512-025-01240-w","url":null,"abstract":"<div><h3>Background</h3><p>As part of the Proryv project, the final stage of reprocessing spent nitride fuel involves the separation of americium and curium on a cation-exchange sorbent using displacement complexation chromatography. For radioecological safety of the environment, spent radioactive sorbent should be converted into a stable compound.</p><h3>Aim</h3><p>To test magnesium potassium phosphate (MPP) matrix for immobilization of spent radioactive sorbent formed during the reprocessing of spent nitride fuel, as well as to determine the quality indicators of the resulting MPP compound.</p><h3>Materials and methods</h3><p>The phase composition of obtained samples was determined by X‑ray diffractometry. Water resistance, mechanical compressive strength, resistance to thermal freeze-thaw cycles, as well as leaching of matrix-forming elements and ^{241, 243}Am and ²⁴⁴Cm isotopes was determined in accordance with standard tests.</p><h3>Results and discussion</h3><p>The obtained samples of MPP compound contain up to 20 wt % of the spent sorbent simulant and 10 wt % of wollastonite; their main crystalline phase represents an analogue of the K‑struvite natural mineral. The compressive strength of MPP compound is ~6 MPa, including upon the completion of freeze-thaw and water resistance tests after 90 days of immersion in water. The compound is demonstrated highly resistant to leaching of matrix-forming elements and isotopes: the leaching rate of ^{241, 243}Am and ²⁴⁴Cm is approximately 5.8∙10⁻⁷ and 2.5∙10⁻⁷ g/(cm²∙day), respectively.</p><h3>Conclusion</h3><p>The prospects of using MPP compound for immobilization of spent radioactive sorbent formed during reprocessing of spent nitride fuel are shown. Quality indicators of obtained samples meet the requirements for solidified radioactive waste.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"154 - 161"},"PeriodicalIF":0.3,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011780","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical study of uranium nitride dissociation time at various temperatures using the EVKLID/V2 code","authors":"E. V. Usov,&nbsp;R. E. Ivanov,&nbsp;V. I. Chukhno,&nbsp;A. A. Butov,&nbsp;I. G. Kudashov,&nbsp;V. D. Ozrin,&nbsp;N. A. Mosunova,&nbsp;V. F. Strizhov","doi":"10.1007/s10512-025-01237-5","DOIUrl":"10.1007/s10512-025-01237-5","url":null,"abstract":"<div><h3>Background</h3><p>Dense nitride fuel is promising for reactors on fast neutrons (fast reactors) with liquid-metal coolant to solve the problems of closing the nuclear fuel cycle. An important aspect of research involves the study of the nitride fuel dissociation rate under various conditions, including that characteristic of severe accidents.</p><h3>Aim</h3><p>To numerically study the time of uranium nitride fuel dissociation under various external conditions characteristic of severe accidents with core destruction.</p><h3>Materials and methods</h3><p>The object of the study is uranium nitride. The research method is numerical simulation using the validated SAFR severe accident module of the EVKLID/V2 integral code developed by the IBRAE RAS, Afrikantov OKBM JSC, NIKIET JSC, and NRC Kurchatov Institute (Russian Federation). The SAFR module uses a model for calculating the behavior of nitride fuel elements in severe accidents. Computational methods of the SAFR module include enthalpy formulation to solve the thermal conductivity equation; the melt flow is simulated by solving one-dimensional equations of mass, energy, and momentum conservation.</p><h3>Results</h3><p>Depending on the properties of the contact environment, time of fuel dissociation may vary by several orders of magnitude. In a leaking fuel element with a helium atmosphere without nitrogen at a relatively high temperature of 1800 ℃, the dissociation time for 50% of uranium nitride ranges from 40 days to ~3 years, depending on the conditions of wetting the fuel column by liquid uranium. At high values, close to the nitride melting point of 2600 ℃, the dissociation time is from hundreds of seconds to 1 h. The presence of nitrogen with a partial pressure of ~2.5 atm completely suppresses dissociation up to melting.</p><h3>Conclusion</h3><p>The longest time of uranium nitride dissociation, and therefore its greatest thermochemical stability, is observed during dissociation into a gas atmosphere in the presence of a residual melt film on the fuel surface.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"133 - 138"},"PeriodicalIF":0.3,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011779","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Current status of the BN-1200M project: a review","authors":"A. V. Vasyaev,&nbsp;A. V. Kerekesha,&nbsp;A. N. Kryukov,&nbsp;E. V. Marova,&nbsp;S. F. Shepelev,&nbsp;A. V. Andreev,&nbsp;A. M. Dyagilev,&nbsp;S. V. Egorov,&nbsp;A. V. Yashkin","doi":"10.1007/s10512-025-01236-6","DOIUrl":"10.1007/s10512-025-01236-6","url":null,"abstract":"<div><h3>Background</h3><p>Development of a two-component nuclear power system with a closed nuclear fuel cycle based on new-generation fast reactors represents an important task put forward in the Strategy for the Development of Nuclear Power in the Russian Federation until 2050.</p><h3>Aim</h3><p>To generalize the results of developments for the project of a new generation commercial power unit with a BN-1200M reactor and to substantiate innovative engineering solutions for its equipment and systems.</p><h3>Materials and methods</h3><p>The BN-1200M project uses both reference engineering solutions that have proven themselves well during the operation of BN-600 and BN-800 power units and advanced solutions for the core, main circulation pumps, steam generator, cold traps, reloading system, and layout of the primary and secondary circuit equipment. The current work is focused on the justification of mixed uranium-plutonium fuel, design of the emergency heat removal system, operational safety, manufacturing technology for structural elements of a two-module vessel steam generator, new structural materials, etc.</p><h3>Results</h3><p>An engineering design for a sodium-cooled BN-1200M fast neutron reactor has been developed; the competitiveness of the project has been substantiated; design materials for the power unit have been developed. The results of R&amp;D on selecting emergency core protection using passive equipment in the emergency heat removal system for natural circulation along the circuits and justifying new structural materials in heat exchange equipment have increased the level of safety, reliability, and cost-effectiveness of the BN-1200M reactor as compared to previous projects and promising energy sources.</p><h3>Conclusion</h3><p>The completed work became a transition to the commercial development of the designed power unit with a high-power BN-1200M reactor plant in a two-component nuclear power system. This stage of commercial development solves the urgent task of completing the experimental validation of engineering solutions for equipment, including the development and testing of prototypes on sodium benches.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"121 - 132"},"PeriodicalIF":0.3,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011781","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 1","authors":"I. M. Russkikh,&nbsp;E. N. Seleznev,&nbsp;A. A. Zyryanova,&nbsp;Yu. V. Volchikhina,&nbsp;N. M. Aristov,&nbsp;N. S. Kalashnikov,&nbsp;A. V. Goryachikh,&nbsp;O. A. Kravtsova,&nbsp;O. L. Tashlykov","doi":"10.1007/s10512-025-01238-4","DOIUrl":"10.1007/s10512-025-01238-4","url":null,"abstract":"<div><h3>Background</h3><p>The main conditions for the safe operation of fuel assemblies (FAs) in the core of the IVV-2M heterogeneous water-water nuclear reactor of the pool type include the absence of surface boiling on the fuel element cladding under the layer of deposits. The values of the coolant flow rate through the FA can be used to predict its temperature at the FA outlet for the campaign and justify the operating limit settings.</p><h3>Aim</h3><p>To determine the analytical dependence of the relative quantitative change in the coolant flow rate through the FA on the pressure drop in the core based on the obtained primary empirical data from hydraulic tests on an FA model.</p><h3>Materials and methods</h3><p>A methodology for measuring the coolant flow rate through FAs is developed. Hydraulic tests of an FA model were performed to determine a formula for calculating the coolant flow rate through the FA in the core depending on its pressure drop.</p><h3>Results</h3><p>The developed methodology is appropriate for determining the coolant flow rate through the FA of the IVV-2M reactor. Measurements carried out on an FA model were used to obtain an empirical dependence of the change in the flow rate on the pressure drop in the core.</p><h3>Conclusion</h3><p>The considered methodology increases safety during reactor operation, as it allows the temperature at the outlet of the FA, and therefore, on the fuel element cladding to be predicted. The analytical formula of obtained dependencies can be used to calculate the coolant flow through the FA in the reactor core and to analyze the measurement results for different core configurations and coolant temperatures. The next article “Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 2” will represent a description of the flow meter design and an analysis of flow rate measurements over several years.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"139 - 144"},"PeriodicalIF":0.3,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011782","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Laser-induced breakdown spectrometry for product monitoring of spent nuclear fuel reprocessing","authors":"A. A. Zherebtsov,&nbsp;Yu. S. Mochalov,&nbsp;M. A. Tarazanova,&nbsp;A. N. Rybakov,&nbsp;K. N. Dvoeglazov,&nbsp;E. A. Shirshin,&nbsp;G. S. Budylin,&nbsp;V. A. Petrov","doi":"10.1007/s10512-025-01239-3","DOIUrl":"10.1007/s10512-025-01239-3","url":null,"abstract":"<div><h3>Background</h3><p>Process monitoring for the composition of spent nuclear fuel (SNF) is closely related to the risks of radiation exposure to personnel and formation of secondary radioactive waste. In addition, it requires labor-intensive sample preparation, as well as costly, long-term, and numerous measurements, making the transition to remote measurement methods, such as laser-induced breakdown spectrometry (LIBS), extremely important.</p><h3>Aim</h3><p>To determine the possibility of introducing the LIBS method into the system for chemical composition monitoring of hydrometallurgical reprocessing of SNF from the BREST-OD-300 reactor plant for an alternative replacement of existing physical and chemical methods including atomic emission spectrometry, spectrophotometry, and X-ray fluorescence analysis for measuring the content of U, Pu, and rare earth elements (REE).</p><h3>Materials and methods</h3><p>A review of literature sources was conducted. A model of the experimental laboratory LIBS unit was developed; test measurements of the element composition were made using the spectral characteristics of solid and liquid samples simulating oxidized SNF.</p><h3>Results</h3><p>The performed literature review showed the potential of using LIBS. The developed LIBS unit was effectively used to determine actinide atomic emission lines using U as an example and identify the presence of nuclear fuel fission products. The measurement error for the concentration of La, Nd, Ce, Fe was 10, 12, 4.5, and 13.5%, respectively, in solid samples of REE oxides with a concentration from 10³ to 10⁵ ppm. For Zr, the error is 3.3% in the samples of nitric acid solutions with a concentration of 0.9–15 g/dm³. Economic efficiency of using the LIBS method involves the reduction in the daily number of measurements and equipment costs by ~29 and 40%, respectively, at the duration of measurements decreased to 24 h.</p><h3>Conclusion</h3><p>The experimental results confirm the operability of the developed LIBS unit and possibility of determining uranium, plutonium, and REE, as well as measuring their content. Due to reduced risks of personnel exposure, the LIBS method can be used to replace a number of quantitative chemical analysis methods for conducting remote online radiation monitoring and decelerating the formation of secondary radioactive waste. In the future, the work on reducing detection limits and measurement errors is required.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"145 - 153"},"PeriodicalIF":0.3,"publicationDate":"2025-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011715","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experience of creating an in situ biogeochemical barrier in contaminated groundwater at nuclear fuel cycle facilities. Part 2","authors":"G. D. Artemyev,&nbsp;A. V. Safonov,&nbsp;I. Yu. Myasnikov,&nbsp;I. E. Kazinskaya,&nbsp;A. P. Novikov","doi":"10.1007/s10512-025-01242-8","DOIUrl":"10.1007/s10512-025-01242-8","url":null,"abstract":"<div><h3>Background</h3><p>Technologies based on the principles of “green chemistry” are actively used around the world to treat contaminated areas. Field studies at radiochemical plants in the USA and China have demonstrated bioremediation potential for complex treatment of groundwater. The article presents the experience of applying this approach in the Russian Federation.</p><h3>Aim</h3><p>To study the mechanisms of immobilizing redox-sensitive radionuclides in the presence of an <i>in situ</i> biogeochemical barrier for treatment of groundwater with polyelement radioactive contamination at the territories of the Siberian Chemical Plant JSC (SCP) and Chepetsk Mechanical Plant JSC (CMP), as well as to justify optimal compositions for the activation of indigenous microflora.</p><h3>Materials and methods</h3><p>The elemental composition of samples was determined using inductively coupled plasma mass spectrometry (ICP-MS); the CGE method of capillary gel electrophoresis was used to determine the ion concentration. Various organic substrates and food production waste were used. Field experiments were conducted on the territories of the CMP (2020), as well as near the SCP B2 pool (2015) and Sublimate Plant (2018).</p><h3>Results</h3><p>Whey, molasses, and a mixture of sucrose and phosphates yielded less than 10% of desorbed actinides. Comparatively high desorption rates are characteristic of stillage, glucose, sucrose, and a mixture of microbial metabolites. The maximum effect is observed in distilled water with 3 g/L of nitrate ions: 33, 55, and 45% desorption for U, Pu, and Np, respectively.</p><h3>Conclusion</h3><p>Field tests have proven high efficiency of using food production waste for the biological treatment of radioactive areas. The biogeochemical barrier zone reduces colloidal and pseudocolloidal transfer of radionuclides due to the coagulation of iron and clay particles. Oxidized iron phases ensure long-term reliable immobilization of redox-sensitive radionuclides under oxidative conditions. The total cost of materials for treating a formation of more than 600 m³ with a single injection of organic mixtures amounted to about 25,000 RUB. The obtained results allow this approach to be recommended for large-scale use to treat the territories near radioactive waste storage reservoirs during operation and post-mothballing periods.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"171 - 179"},"PeriodicalIF":0.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011977","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Method for utilizing contaminated fractions of reprocessed uranium enrichment in double cascade schemes","authors":"Vladislav E. Gusev,&nbsp;Andrei Yu Smirnov,&nbsp;Georgy A. Sulaberidze","doi":"10.1007/s10512-025-01245-5","DOIUrl":"10.1007/s10512-025-01245-5","url":null,"abstract":"<div><h3>Background</h3><p>Increasing the efficiency of using reprocessed uranium can partially solve the problems of resource availability for thermal neutron power reactors, as well as the continuous accumulation of spent nuclear fuel.</p><h3>Aim</h3><p>To develop a method for including the double cascade subproduct in the production of nuclear fuel without additional costs for separative work by replacing part of the natural uranium in producing low-enriched uranium for thermal neutron reactors.</p><h3>Materials and methods</h3><p>The possibility of involving the subproduct of a modified double cascade for enrichment of reprocessed uranium in the production of nuclear fuel is considered. A method is proposed for including the subproduct of such a double cascade in the production of nuclear fuel.</p><h3>Results</h3><p>The study demonstrates this method saving natural uranium by 4.3% compared to the same cascade without using the light fraction flow.</p><h3>Conclusion</h3><p>The proposed method increases the proportion of ²³⁵ U returned to the fuel cycle and provides additional combined savings of natural uranium and costs of separative work compared to an open nuclear fuel cycle.</p></div>","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"197 - 202"},"PeriodicalIF":0.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011981","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Forecast of tritium release into the environment during the normal operation of nuclear power plants","authors":"D. A. Vlasov,&nbsp;D. D. Desyatov,&nbsp;A. A. Ekidin","doi":"10.1007/s10512-025-01241-9","DOIUrl":"10.1007/s10512-025-01241-9","url":null,"abstract":"&lt;div&gt;&lt;h3&gt;Background&lt;/h3&gt;&lt;p&gt;Tritium (&lt;sup&gt;3&lt;/sup&gt;H) is one of the main dose-forming radionuclides in emissions and discharges of nuclear power plants (NPPs). Taking into account the current levels of &lt;sup&gt;3&lt;/sup&gt;H discharges and emissions, its release into the environment can be predicted for the future change in electricity production, as well as during the construction of new promising power units.&lt;/p&gt;&lt;h3&gt;Aim&lt;/h3&gt;&lt;p&gt;To forecast the release of &lt;sup&gt;3&lt;/sup&gt;H from emissions and discharges during normal operation of NPPs with different types of reactors.&lt;/p&gt;&lt;h3&gt;Materials and methods&lt;/h3&gt;&lt;p&gt;A large array of data on the annual discharge and emission of &lt;sup&gt;3&lt;/sup&gt;H by NPPs was analyzed: 66, 10, 55, and 38 NPPs in the USA, Russia, Europe, and other countries, respectively. The normalized values for the annual activity of emissions and discharges to the annual production of electrical energy are taken as specific emission and discharge indicators.&lt;/p&gt;&lt;h3&gt;Results&lt;/h3&gt;&lt;p&gt;An analysis of data on the annual activity of &lt;sup&gt;3&lt;/sup&gt;H released into the environment during normal operation of NPPs with all types of reactor plants was conducted. The dynamics of &lt;sup&gt;3&lt;/sup&gt;H specific emissions and discharges in the period 1985–2022 is considered. The absence of a tendency towards an increase in the indicators was shown for all types of nuclear reactors, with the exception of PWRs and BWRs. The ranking of NPPs with seven types of reactors (PWR, BWR, VVER, RBMK+EGP, GCR+AGR, PHWR, and FBR) was carried out in the best, worst, and sustainable categories according to the value of &lt;sup&gt;3&lt;/sup&gt;H specific emissions and discharges. A forecast of tritium release into the environment from emissions and discharges of NPPs was made under optimistic, probable, and pessimistic scenarios.&lt;/p&gt;&lt;h3&gt;Conclusion&lt;/h3&gt;&lt;p&gt;To assess the compliance of newly designed NPPs with the basic INPRO principle of sustainable development for the nuclear power system, we determined the minimum achievable levels of specific indicators: 2.62∙10&lt;sup&gt;−2&lt;/sup&gt; and 7.19∙10&lt;sup&gt;−2&lt;/sup&gt; GBq/(GWh) for &lt;sup&gt;3&lt;/sup&gt;H emissions and discharges, respectively. According to our forecast of the total tritium release from emissions and discharges of NPPs in 2050, the tritium activity will not exceed 1.38∙10&lt;sup&gt;7&lt;/sup&gt; GBq under the optimistic scenario; pessimistic and probable scenarios yield in 4.73∙10&lt;sup&gt;7&lt;/sup&gt; and 3.73∙10&lt;sup&gt;7&lt;/sup&gt; GBq, respectively. At the upper limit of NPP power generation, the accumulated activity of tritium, taking into account its decay by 2050, is estimated at 4.68∙10&lt;sup&gt;8&lt;/sup&gt; GBq, which will be 18% of its natural equilibrium content. The present work is the 2nd part of a series of articles on tritium: the study uses the results of the 1st part “Global release of tritium into the environment: NPP reactor type contribution”. The next work in the series, planned for publication in the near future, aims to assess individual and collective doses of tritium entering the environment duri","PeriodicalId":480,"journal":{"name":"Atomic Energy","volume":"138 3","pages":"162 - 170"},"PeriodicalIF":0.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145011976","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}