Fission Product Release Research Articles

CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

Investigation of Fission Product Release from Irradiated Oxide and Nitride Nuclear Fuel on High-Temperature Heating in Helium Flow

Investigation of Fission Product Release from Irradiated Oxide and Nitride Nuclear Fuel on High-Temperature Heating in Helium Flow

Experimental study on thermo-chemical behavior of PHWR fuel channel under slumped fuel pin condition

The safe operation of the reactor is compromised under postulated accident events. The events may originate from single or multiple failures. LOCA, along with the unavailability of ECCS, is one such multiple failure event. Phenomena like fuel heat up, fuel slumping, clad oxidation, hydrogen generation and fission product release are about to take place. However, moderator cooling to the external surface of fuel channels limits the fuel heat up. The steam starvation in the channel limits hydrogen generation. The purpose of this study is to analyze the thermo-chemical behavior of a fuel channel of PHWR, assuming that the 37 fuel elements of a bundle are collapsed and collected in a tightly packed stack at the bottom of the pressure tube of the fuel channel. The temperature distribution is obtained under a pseudo-steady-state for the various components of the simulated fuel channel exposed to an oxidative environment. Experimental results have indicated that a significant temperature gradient gets established in radial, circumferential and axial directions in the fuel channel due to slumped fuel bundle configuration as well as hydrogen generation. Estimation from the heat balance studies shows that moderator functions as an active heat sink.

Assessment of Cesium Compound Behavior during Simultaneous Failure of Reactor Pressure Vessels and Spent Fuel Pools Using Modified ART Mod 2: Fukushima Daiichi Accident Simulation

To support the regional strategy development of ASEAN NPSR using scientific research, Modified ART Mod 2 has been used to assess the fission product release from RPVs and SFPs independently. However, the Fukushima Accident suggested the possibility of simultaneous release from RPV and SFP which indicated the necessity of re-evaluation of the maximum source term. The objective was to assess the fission product behavior during a simultaneous failure in RPVs and SFPs of BWR type with Mark I containment design in multiple units using Modified ART Mod 2 in order to evaluate the maximum source term. The releases of cesium compounds in gas and aerosol forms from RPVs and SFPs of Units 1–3 of the Fukushima Daiichi NPP were selected as the case studies. It was found that the behavior of cesium compounds was mainly governed by the aerosols and atmospheric temperatures, which resulted in different characteristics in adsorption and thermophoresis. It also turned out that the simulation of a simultaneous release led to a smaller release than the summation of independent simulations of releases from RPV and SFP by 25%. This study helped estimate the maximum consequences in order to be able to effectively design the EPR for NPP accidents inside or outside the ASEAN region.

Fuel integrity evaluation method for BWR-6 plant Kuosheng during ATWS events

Abstract Anticipated Transients without Scram (ATWS) are evaluated to demonstrate that during ATWS the fuel’s integrity prevents fission product release, and that the safety features the BWR-6 plant is equipped with mitigate the ATWS event. For analyzing this scenario, a RETRAN-3D system model and hot channel model of Kuosheng NPP have been developed. From all possible events those are analyzed where the main steam isolation valves close (MSIVC) when the reactor is operating at rated power conditions. For cases with low core flows this results in higher peak cladding temperatures and thicker oxidation thicknesses. All results confirm the ATWS acceptance criteria. For sensitivity cases, increasing the boron concentration of standby liquid control system (SLCS) or injecting boron of SLCS earlier were investigated. The results show, that these measures cannot decrease the PCT value immediately, but can shorten the period of peak cladding temperature over 850 °C which results in smaller oxidation thickness.

Effect analysis of ISLOCA pathways on fission product release at Westinghouse 2-loop PWR using MELCOR

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1–4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

Experimental study of SRT scrubbing model in water coolant pool

In pool-type nuclear reactors, a postulated fuel pin failure can result in the release of fission products into the coolant pool. While the coolant pool itself tends to scrub the majority of radionuclides, they have the potential to be transported to the cover gas region via entrainment in gaseous bubbles. The SRT source term analysis code includes models to assess this pool scrubbing behavior and calculate the ratio of fission product aerosols that reach the cover gas. This code was developed to model pool-type sodium cooled fast reactors, but underlying models can also be used to predict pool scrubbing in a water coolant pool with dry air bubbles passing through as a test of the physics. The accuracy of this code is analyzed in this paper. A parametric study was conducted to determine the significance of bubble size, aerosol density, pool depth, and aerosol concentration, over a wide range of aerosol sizes, on the pool scrubbing efficiency. The basic tests were run with isolated bubbles as to best match the assumptions within the code. For simulating potentially realistic pool scrubbing conditions, an applied test was run injecting a bubble swarm into the pool to analyze the bubble-to-bubble interactions. From these results, it was found that the SRT code matches the trends found experimentally, but underestimates the actual amount of pool scrubbing. The bubble swarm testing further exemplifies this trend. Furthermore, it was determined that the overall pool scrubbing increases with bubble size, aerosol density, and pool height and is unchanged by aerosol concentrations. Additionally, aerosol sizes ranging from one to 0.01 µm were found to be of particular interest as the lowest amount of scrubbing occur over this range. These tests also inform the design of the experiment which will be conducted in sodium.

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2 – Results of severe accident analyses for unit 3

This is the third part of the three part paper describing the accidents at the Fukushima Daiichi nuclear power station as analyzed in the Phase 2 of the OECD/NEA project “Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant” (BSAF). In this paper, we describe the accident progression in unit 3. Units 1 and 2 are discussed in parts 1 and 2 of this series of papers.In the BSAF project, eight organizations from five countries (CRIEPI, IAE, JAEA and NRA, Japan; IRSN France; PSI, Switzerland; SNL, USA; VTT, Finland) analyzed severe accident scenarios for Unit 3 at the Fukushima Daiichi site using different severe accident codes (ASTEC, MAAP, MELCOR, SAMPSON, THALES). The present paper for Unit 3 describes the findings of the comparison of the participants’ results against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on the status of the reactor pressure vessel, melt release and fission product release and transport. Unit 3 specific aspects, e.g., the complicated accident progression following repeated containment venting actuations and attempts at coolant injection at the time of the major core degradation, are highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized. Fission product transport is analyzed, and the calculation results are compared with dose rate measurements in the containment. The release of I-131 and Cs-137 to the environment is compared with analysis conducted using WSPEEDI code.

Modeling fission product diffusion in TRISO fuel particles with BISON

The diffusion of fission products in intact TRISO particles depends on particle geometry, fission product source rates, time, temperature, and temperature-dependent diffusion coefficients. Simulating this diffusion process requires models for source rates and diffusion coefficients, plus a computation of the temperature field if not prescribed. In addition, the simulation quality depends on the discretization of the geometry, appropriate time stepping, and the accuracy of the solution method.In this paper, we explore the simulation of fission product diffusion in TRISO fuel particles using the finite element method via the fuel performance code BISON. Recent material model development has occurred in BISON for each material present in tri-structural isotropic (TRISO) fuel particles: the buffer, inner pyrolytic carbon, silicon carbide, and outer pyrolytic carbon layers, as well as the fuel kernel. Also, new mesh generation and fission product release fraction capabilities have been added.Diffusion capabilities are shown to converge to the correct solution via formal verification tests. A large number of code benchmarking problems are also given, with good results, showing that BISON’s computed release fractions closely match those of other software tools. Finally, a significant validation effort is detailed in which fission product release, measured as part of the AGR-1 capsule experiments, is compared to BISON outputs. BISON outputs compare favorably to the experimental data and to PARFUME results.

A Model of the Fission Products Release from the Melt Pool during Severe Accident in a Liquid Metal Cooled Fast Reactor

Accident scenarios implying the failure of the active and passive reactor protection system are the most severe in their consequences since they are accompanied by the melting of fuel elements, fuel assemblies, and the entire core and, as a consequence, the release of fission products (FP) from the fuel melt and structural materials. The location of reactor facilities near populated areas requires an understanding of the consequences of hypothetical accidents with the destruction of the core. The most important is information on the amount of FP released into the environment. To assess the possible dose load on personnel and the environment in a hypothetical accident, data on the release of radioactive nuclides from the damaged, including molten, core are required. This work presents approaches that can be used to calculate the amount of fission products released from the fuel melt. The proposed approaches are based on general patterns of behavior and release of impurities from liquid solutions. To simulate this process, a system of differential equations is solved. The proposed approaches were implemented in the form of a software module for calculating the release of FP from the melt. Using the developed module, a test problem for the release of FP from a molten oxide fuel was solved, and an analysis of the results of real experiments carried out at the National Laboratory in Oak Ridge was carried out. Satisfactory convergence of the calculated data on the yield of both nonvolatile (Ce, Sr) and volatile (Xe, Kr) fission products is shown.

Preliminary Study of DLOFC Accident in HTGR using GRSAC

High Temperature Gas-cooled Reactor (HTGR) is one of the Generation-IV reactors that has many advantages and very promising to be utilized in the future. National Nuclear Energy Agency of Indonesia (BATAN) has a program to build a non-commercial power reactor (RDE) that will be based on the pebble-bed type of HTGR. To support this program, the research and development activities in HTGR have been conducting recently. One of the important aspects to be studied is accident analysis to assess the behavior of HTGR during the accident and to anticipate the consequences that might be caused. The purpose of this work is to study the Depressurized Loss of Force Coolant (DLOFC) accident on HTGR by using Graphite Reactor Severe Accident Code (GRSAC). The type of HTGR used as a case study is PBMR-400. This work was begun with the preparation of GRSAC input for reactor parameter including reactor design, core layout design, primary cooling loop design, graphite properties and material properties, cavity design, vessel design, and fission product release. It was assumed that there is an initial event that causes the reactor depressurization and the coolant loss during the time range of 13 seconds. During this time, the reactor pressure dropped from 91.75 kg/cm2 to 1.05 kg/cm2 and the coolant flow reduced from 192.53 kg/s to 0.29 kg/s. Two types of DLOFC event were studied which are DLOFC with scram and DLOFC without scram. The behavior of the fuel temperature was then analyzed during the time range of 100 hours after the DLOFC accident. It was obtained that the maximum fuel temperature was reached at the time of 60 hours after the accident. The maximum temperature after the DLOFC with and without scram accident is 1545oC and 1740oC, respectively. The results of this work were also compared to the results of benchmark work on PBMR-400 conducted by American Nuclear Society and it showed a good comparison. At the end, this work will be very beneficial to be implemented in the RDE program.

Gas detection in sodium cooled fast reactors: determination of a transfer function

To keep the dependability of Sodium Cooled Fast Reactor, the “clean sodium concept” is demanded, which means that the sodium is free from contamination. The release of fission products is searched for by a contamination measuring system. We need to have a comprehensive description of cladding failures and the detection of contamination, including the failure occurrence on the fuel pin, the transfer process through the sodium and cover gas, the measurement efficiency, etc. We aim to identify the important parameters of physical phenomena, with modelling and simulations based on the return of experiments from past reactors such as PHENIX. There have been a total of 15 open pin failures in PHENIX reactor. Through studying these detected signals, we can get a better physical explanation and description of the evolution of failures. The detection system is related to different stages of the evolution of fuel pin, with different types of fission products, various release mechanism and physical properties. During the evolution of the failed fuel pin, gaseous fission products is released on the first stage of failure and the gas detection system is aimed for the gaseous fission products monitoring. We proposed a quantitative modeling of transfer function to describe the time broadening of the gas release from the fuel pin to the detector. The result matches well with PHENIX experiment data, with the same order of magnitude of the time broadening and the same shape of exponential decreasing. A determination of the transfer function of this gas detection system is validated in this paper.

ASSESSMENT OF HEAT TRANSFER IN FUEL CHANNEL OF INDIAN PHWR UNDER POSTULATED LARGE BREAK LOSS OF COOLANT ACCIDENT: EXPERIMENTAL AND NUMERICAL STUDY

The behaviour of the channel under postulated large break LOCA scenario had been a prime safety concern. The radiative heat transfer is predominant in a channel when the convective cooling environment is marred. The estimation of temperature distribution in the fuel pins at elevated temperature is essential from the point of view of hydrogen gas generation and release of fission products. In this paper, the thermal characteristics of a channel for Indian PHWR under critical break failure is studied using experimental and numerical techniques. The experiment is carried out on an Indian PHWR having a fuel bundle of 37-fuel elements. The temperature profiles for different parts of the simulated channel comprising of fuel pins, PT and CT are obtained under steady condition. The numerical analysis is also performed using ANSYS Fluent 19.0. From the study, it is found that there is a significant radial temperature gradient in the fuel bundle from the center ring to the outer ring. Also, no significant circumferential temperature gradient is observed in the fuel bundle, PT and CT.

Microstructure of zirconium fuel claddings: TEM and EBSD studies of as-received and ne-utron-irradiated materials

Zirconium fuel claddings act as a first barrier against release of fission products during nuclear power plant operation and interim storage of the spent fuel. During the reactor operation, cladding tubes are exposed to different stress level at elevated temperatures and neutron irradiation in corrosive environment. It causes a material degradation by corrosion, cladding embrittlement by hydrides and radiation-induced damage or radi-ation growth and creep of the fuel rods. The irradiation damage effects mainly contribute to the loss of material ductility. In our study, microstructure of as-received (non-irradiated) Zr-alloys used in LWR (Zr1Nb, Zr-1Nb-1.2Sn-0.1Fe, Zr-1.5Sn-0.2Fe-0.1Cr) were examined by electron microscopy methods. Transmission electron microscope (TEM) was used to describe the microstructure of claddings used in different reactor conditions and identify the radiation-induced damage, which is presented on Zr1Nb irradiated to one standard campaign in the VVER-1000 active zone. Following Electron Backscatter Diffraction (EBSD) method on transparent foils complements the TEM results in larger area, i. e. by grain size and orientation or analysis of local misorienta-tion after irradiation. Radiation-induced damage was observed in Zr1Nb metallic matrix as type disloca-tion loops, presence of radiation-induced precipitates or partial amorphization of the secondary phase particles. EBSD method showed no changes in crystallographic orientation, but a local increase of dislocation density can be affected by neutron irradiation.

Estimation of fission product source terms for the SGTR accident of a reference PWR plant using MELCOR and MAAP5

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

Modelling of fission products release in VERDON-1 experiment with cGEMS: Coupling of severe accident code MELCOR with GEMS thermodynamic modelling package

The treatment of chemistry and thermodynamics in the integral severe accident codes is typically limited. A more accurate treatment of the chemistry during the severe accident modelling is, therefore, of great interest. For this purpose, the work is focused on the development of coupling of the MELCOR code with chemical thermodynamic calculations using GEMS codes and HERACLES database. Developed coupling between the two codes, called cGEMS, allowed for the improved thermodynamic description of the fission product release from the nuclear fuel under severe accident conditions.VERDON-1 test was selected as an experimental reference for the simulations. Experimental release behaviour of Mo, Cs and Ba observed in VERDON-1 experiment was reproduced by the developed coupled code. Performed simulations provided detailed information about the fission product speciation at different redox conditions.The obtained information and the developed code provides a more accurate description of the fission product behaviour and release during severe accidents.

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

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.

Estimation of fission product source terms for the ISLOCA of a reference PWR plant using MELCOR and MAAP5

An interfacing systems loss of coolant accident (ISLOCA) is a type of containment bypass accident that may lead to a significant release of fission products (FPs) to the environment. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during the ISLOCA of a reference pressurized water reactor (PWR) plant and investigates the influence of a dedicated accident mitigation action on the results of interest through the established modeling approach. In addition to this, the key plant features and phenomenological mechanisms that might have a significant influence on the magnitude and timing of FP releases to the environment during an ISLOCA were examined through a couple of sensitivity analyses. To consider the integral responses of the reactor, containment, and auxiliary building to a severe accident initiated by an ISLOCA, detailed modeling was performed for a broad spectrum of relevant phenomena, plant systems, and structures that promote the deposition and retention of FPs along the flow path to the environment. As the modeling and simulation tools for the ISLOCA analysis, MELCOR 2.2 and MAAP5.04 were employed to compare the results of interest. The results of the analysis shows that the present best practice modeling approach proposed here has the potential to significantly decrease the release of FPs into the environment in the event of an ISLOCA. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), phase 2 – Results of severe accident analyses for unit 2

This is the second paper in a series of 3 in which results of severe accident analyses for Unit 2 of Fukushima Daiichi are presented, gained in phase 2 of the OECD/NEA project “Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)”. Nine organizations of six countries (CNL Canada; CRIEPI, IAE, JAEA and NRA Japan; IRSN France; KAERI Korea; NRC/DOE/SNL U.S. and VTT Finland) submitted results of their calculated severe accident scenarios for Unit 2 of Fukushima Daiichi using different severe accident codes (ASTEC, MAAP, MELCOR, SAMPSON, THALES-KICHE).The present paper describes the findings of the comparison of the participants’ results for Unit 2 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on reactor pressure vessel status, melt release and fission product behavior and release. Unit 2 specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized.

Raman spectroscopy coupled to principal component analysis for studying UO2 nuclear fuels with different grain sizes due to the chromia addition

Current necessities of nuclear power plants have led to increase burn-up of the fuel during operation. In this context, some undesirable processes, such as a higher release of gaseous fission products, may occur. New UO2-based fuels are being developed by adding doping agents. Chromia (Cr2O3) has proved to considerably enhance grain growth during sintering, and thus promote the retention of these fission products. The study of these Cr-doped UO2 fuels at in situ conditions would allow testing the real performance of the fuel in operating conditions. The existence of Raman portable instruments makes this scenario feasible. But first, the measurement protocols need to be developed. Therefore, in this paper, we explore the use of Raman spectroscopy as an approach for the analysis of the effect of Cr2O3 addition in a set of UO2Cr2O3 sintered pellets. The validity of Raman is demonstrated by using Principal Component Analysis (PCA). Three Principal Components describe 98.8% of the total variance of the data, and they are related to the main Raman modes of the samples. In addition, SEM images have shown the presence of bigger precipitates of Cr2O3 not only when the solubility limit of Cr3+ is exceeded, but also when it is not reached. By XRD, the well-known Vegard behavior is observed, and a solubility limit of Cr3+dissolved into the UO2 matrix is found to be (748±16) ppm for the particular sintering conditions used.