Optimization Study on LLFPs Transmutation in PWRs

2013 ◽  
Author(s):  
Kun Liu ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Youqi Zheng
Keyword(s):  
Systematic Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Zirconium Hydride ◽  
The Core ◽  
Optimization Study ◽  
Safety Parameters ◽  
Negative Impacts ◽  
Isotopes Separation

A systematic study on the Long-lived fission products (LLFPs) transmutation in a PWR has been performed, aiming to devise optimal transmutation strategy in present nuclear power plants. The LLFPs selected in the analysis include 99Tc and 129I discharged from LWRs. The isotope, 127I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, technetium or MgI2 targets mixed with zirconium hydride are designed and investigated. The equilibrium cycles are investigated. The transmuted amounts of 99Tc and 129I are equals to the yields from 1.94 and 4.22 1000 MWe PWRs, respectively. Numerical results indicate that both the 99Tc and 129I can be transmuted conveniently in present PWRs in the form of target pins.

Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS

2006 ◽  
Author(s):  
Ingo D. Kleinhietpaß ◽  
Hermann Unger ◽  
Hermann-Josef Wagner ◽  
Marco K. Koch
Keyword(s):  
Phase Boundary ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Energy Systems ◽  
The Core ◽  
Gas Atmosphere ◽  
Upper Pool

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport.

scholarly journals Development of a criterion for assessment of fuel washout during operation of WWER power units

2020 ◽  
Vol 6 (4) ◽  
pp. 307-312
Author(s):  
Igor A. Evdokimov ◽  
Andrey G. Khromov ◽  
Petr M. Kalinichev ◽  
Vladimir V. Likhanskii ◽  
Aleksey A. Kovalishin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Reactor Operation ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Reliable Detection ◽  
Coolant Activity

Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

scholarly journals Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽  
2019 ◽  
Vol 6 (11) ◽  
pp. 120
Author(s):  
Luca Albertone ◽  
Massimo Altavilla ◽  
Manuela Marga ◽  
Laura Porzio ◽  
Giuseppe Tozzi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Scaling Factors ◽  
Activation Products ◽  
Long Time ◽  
Nuclear Installation ◽  
Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

The Effect of Elevated Temperatures and Nuclear Radiation on the Properties of Biological Shielding Concrete

2016 ◽  
Vol 677 ◽  
pp. 8-16 ◽  
Author(s):  
Jaroslava Koťátková ◽  
Jan Zatloukal ◽  
Pavel Reiterman ◽  
Jan Patera ◽  
Zbyněk Hlaváč ◽  
...  
Keyword(s):  
Gamma Radiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elevated Temperatures ◽  
Nuclear Radiation ◽  
Containment Vessel ◽  
Negative Impacts

The paper reviews the so far known information about the properties of biological shielding concrete used in the containment vessel of nuclear power plants (NPP) and its behaviour when exposed to radiation. The damage of concrete caused by neutron and gamma radiation as well as by the accompanying generation of heat is described. However, there is not enough data for the proper evaluation of the negative impacts and further research is needed.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

scholarly journals HEALTH IN THE SYSTEM OF VALUES OF THE POPULATION OF THE SURVEILLANCE ZONES NUCLEAR POWER PLANTSS

2021 ◽  
Vol 26 ◽  
pp. 219-234
Author(s):  
V. Prylypko ◽  
◽  
Yu. Ozerova ◽  
I. Bondarenko ◽  
M. Morozova ◽  
...  
Keyword(s):  
Family Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sampling Error ◽  
Close Correlation ◽  
The State ◽  
The Core ◽  
Terminal Values ◽  
Happy Family

Objective: to determine the place of health in the system of values of the population of the surveillance zone (SZ) of nuclear power plants (NPPs) and its importance in the perception of emergency risks (ER). Materials and methods. To determine the place of health in the value system, a survey of the able-bodied population of satellite cities of Rivne (RNPP) and South Ukrainian (SUNPP) nuclear power plants was conducted using nonrepetitive sampling, where the sampling error does not exceed 7,0 %. The motivational and behavioral component that determined health in the individual hierarchy of values of the subject according to the questionnaire Berezovskaya R. A. was studied. Statistical and mathematical methods were used in the research process. Results. The array of respondents was conditionally divided into 4 groups according to their attitude to human health. And the group where a person’s life position is focused exclusively on health is the most common – 77,0 %. Group IV, which wants to live without limiting itself, is 8,1 %. The component integrity of values-goals and valuesmeans among the urban population of the SZ of both nuclear power plants is the same: the main goal in life is health, happy family life, and as a means – perseverance, diligence and health. Goal values in groups I and IV have some differences: in the first group of respondents the main goal in life is health, and in the fourth, where a person’s life guidelines exclude any restrictions – a happy family life. Values for these populations have some differences, but in both groups health appears to be the main means to an end. There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Conclusions. Identified hierarchy of values: a group of stable dominant values; average status values; group of least significant values. The values of the highest status among the values-goals are – health, happy family life and interesting work. Most respondents plan to achieve them through values such as «health», «perseverance and hard work». There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Key words: health, values, population, NPP surveillance zone, perception of emergency risks.

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