Water Ingress Accident Research Articles

A proposed refueling scheme for the modular pebble bed reactor

In this paper, a suggested design of the fuel pebble and loading pattern in the MIT Modular Pebble Bed HTGR (MPBR) using thorium (232Th) is presented. Using the MCNP5 code model for the MPBR reactor, different compositions of Thorium Low Enriched Uranium (ThLEU) with different Heavy Metal (HM) loading and 235U enrichment were investigated to achieve the maximum thorium utilization with relatively minimum reactivity insertion in case of water ingress accident. Using pure Low Enriched Uranium (LEU) pebbles and ThLEU pebbles with the same HM loading and 235U enrichment, different pebbles loading patterns were studied. One loading pattern was selected to achieve the criticality with enough excess reactivity compensating for the poisoning effect of xenon build-up at the startup of the reactor operation and the negative effect of fuel and reflector temperature coefficients of reactivity. This loading pattern aims also to reduce the radial power peaking in the core and to minimize the axial Power Peaking Factor (PPF) in the central hot channel, with a minimum number of pure LEU fuel batches. A calculation procedure was developed that combines fuel pebble depletion and full core calculations to investigate the impact of a proposed online refueling scheme for the MPBR reactor and to simulate the multicycle refueling mechanism. Using SERPENT 1.1.7, the nuclide concentrations were calculated as a function of pebble burnup for both LEU and ThLEU pebbles. A multicycle online refueling scheme was proposed and the average number of passes required for LEU and ThLEU pebbles to reach their maximum burnup was estimated.

Transport and deposition behaviors of graphite aerosol in the steam generator building during an overpressure discharge accident of HTR-PM

The water-ingress accident is one of design basis accidents of high-temperature gas-cooled reactors (HTGR). The rapid change of the flow field and the direct brush of the high-velocity gas flow may cause a large fraction of graphite dust to resuspend, which is further discharged into the steam generator building when the overpressure relief protection is triggered. Due to the coupling with the radioactive fission products (FPs), the behaviors of the discharged particles are crucial for the reactor safety analysis and source-term evaluation in the overpressure discharge scenario. In this work, we numerically investigate the transport, deposition and releasing behaviors of graphite aerosol in the full-scale steam generator building of HTR-PM by the Eulerian-Lagrangian method. The mixture-ideal-gas law is used for describing the mixture between helium and air, and the CFD-DPM scheme is employed for tracking the trajectory and velocity of individual particles. Particularly, the effects of the irregular shape on the particle-fluid and particle-wall interactions are included. The simulation results show the particles are quasi-uniformly distributed in the steam generator building and the deposition and releasing rates with the time follow the exponential relationship at the late-stage. The distribution of all the deposited, escaped and airborne particles are analyzed. Due to the strong thermophoretic effect, a large fraction of particles are deposited on the cold walls. The extrapolation by the exponential relation indicates a final releasing fraction of 18.94%. Compared to the current conservative assumption that all the dust particles in the steam generator building will eventually enter the environments, our results demonstrate that the steam generator building of HTR-PM still has certain capability to retain the graphite dust during an overpressure discharge accident scenario.

A CAST-Based Analysis of the Metro Accident That Was Triggered by the Zhengzhou Heavy Rainstorm Disaster.

Emergency management research is used to deal with the increasing number of extreme weather threats in urban areas. This paper uses causal analysis based on systems theory (CAST) to review the subway water ingress accident and the government’s emergency management actions in Zhengzhou, Henan Province, during the heavy rainstorm disaster on 20 July 2021. The aims of this article are to establish safety control structures at both the enterprise level and the government level, and to systematically analyze the problems in emergency management in Zhengzhou City. Our analysis found that the construction of disaster prevention facilities restricted emergency management. Therefore, we suggest that enterprises and governments not only pay attention to emergency management, but also to the construction of disaster prevention facilities. This article also points out that the system of chief executive responsibility that is implemented in China is becoming a double-edged sword in emergency management. Our study makes recommendations for enhancing the capacities of emergency management, points out the shortcomings of the existing emergency management structure, and provides knowledge gained for future emergency management research.

ASSESSMENT OF RADIOLOGICAL IMPACTS FROM POSTULATED ACCIDENT CONDITIONS OF HTGR: A CASE STUDY IN SERPONG NUCLEAR AREA

High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRISO coated fuel particles that are considered not to be damaged even in accident condition. However, the radiological impacts from accident condition in HTGR is still important to be assessed. This research is aimed to perform radiological impacts assessment of two postulated accidents of HTGR, which are depressurization and water ingress accident. As a case study, a 10-MWTh pebble-bed HTGR design named Reaktor Daya Eksperimental with the planned site located in Serpong Nuclear Area was chosen. The source terms from the accident conditions were estimated using mechanistic source term model and the dose consequences were calculated using PC-COSYMA. The input data for PC COSYMA, which are meteorological, population distribution, agricultural and local farm data, were compiled based on the site data of Serpong Nuclear Area. The radiological impacts were assessed based on individual and collective doses. The results showed that the highest dose will be received by the community within a radius of 250 m to the south from the reactor. It was also found that these accidents only cause minor radiological impacts which do not meet the criteria for any countermeasures (iodine thyroid blocking, sheltering, evacuation, food ban, decontamination, and relocation).

Analysis of water ingress accident in the RDE core using MCNPX

The 10 MW Indonesia’s Reaktor Daya Eksperimental (RDE) is a Pebble Bed High Temperature Reactor-type and to be constructed in Serpong Nuclear Zone, Puspiptek, Indonesia. The reactor applies helium gas coolant, graphite moderator, and low uranium enriched fuel (17%). A water-ingress accident may take place in the RDE core and such the accident is caused by a pipe break in the steam generator system of the reactor and hence the accident influences the reactivity of the core. To begin with, an initial core of the RDE should be modelled and the model contains the identification of material zones in the reactor core. To estimate the reactivity change due to the water-ingress accident in the initial RDE core, a void core condition from any coolant should be firstly deliberated. The core reactivity change is then estimated based on effective multiplication factor (k eff) as a function of water density which comes into void cells among pebble containing helium coolant in the RDE core and that of the variation of 235U enrichment. The data library of thermal scattering S(α,β) as a function of temperature has been applied. The data base of thermal scattering for graphite was also utilized to comprehend the validity calculation using MCNPX. The ENDF/B-VII continuous energy nuclear data library was also smeared for all calculations at temperature of 900 0K. For the calculation using MCNPX, two options of KCODE and KSRC were taken into account to estimate k eff of the RDE core. To achieve the accuracy of keff in the RDE reactor core, the amount of 20,000 neutrons per cycle for 150 non-active cycles and 650 active cycles was taken into account. The results showed that water-ingress accident in the RDE core is a dynamical behaviour which may result in reactivity core change of the RDE core. Finally, from all caclulated results taking into account the varietes of water density and 235U fuel enrichment, the RDE reactor core is totally safe in the event of water-ingress accident ensuing during the RDE reactor operation.

ESTIMATION OF THE RADIOACTIVE SOURCE TERM FROM RDE ACCIDENT POSTULATION

The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carried out by BATAN for the last five years, adopting HTGR-type reactor with thermal power of 10 MW. RDE is designed with the reference of similar reactor, namely HTR-10. During this process, source term estimation is required to prove the safety of RDE design, as well as to fulfill the concept of As Low As Reasonably Achievable (ALARA) in radiation protection. The source term is affected by the magnitude of the radioactive substances released from the reactor core due to an accident. Conservative accident postulations on the RDE are water ingress and depressurization accidents. Based on these postulations, source term estimation was performed. It follows the mechanistic source term flow, with conservative assumptions for the radioactive release of fuel into the coolant, reactor building, and finally discharged into the environment. Assumptions for the calculation are taken from conservative removable parameters.The result of source term calculation due to the water ingress accident for Xe-133 noble gas is 8.97E+12 Bq, Cs-137 is 3.59E+07 Bq, and I-131 is 4.34E+10 Bq. As for depressurization accident, the source term activity for Xe-133 is 3.90E+13Bq, Cs-137 is 1.56E+07 Bq, and I-131 is 1.89E+10Bq. The source term calculation results obtained in this work shows a higher number compared to the HTR-10 source term used as a reference. The difference is possibly due to the differences in reactor inventory calculations and the more conservative assumptions for source term calculation.Keywords: RDE, HTGR, Radioactive, Source term, accident

Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach

The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

Estimation of radioactivity impact for RDE based on HTR-10 hypothetical accident - a case study

BATAN priority activity supported by the Center for Nuclear Reactor Technology and Safety is the conceptual design documents and evaluation of experimental power reactor (RDE). Based on the design established, the radiation safety involving the dispersion of radioactive into the site and environment shall be calculated. The objective of this study is to obtain the radioactivity impact of RDE due to hypothetical accident in Serpong II Nuclear Area (KNS-II) in Puspiptek Area. Postulated hypothetical accidents are water ingress and depressurization accident. The sourceterm input data taken are based on HTR-10 hypothetical accident. Meteorological and environmental data are taken from available data of KNS Serpong. The calculation is carried out using PC-Cosyma code. The highest radioactivity of air dispersion for Kr-87due to depressurization accident is 1.96E+04 Bq/m3 and due to water ingress accident is 1.96E+04 Bq/m3. The highest radioactivity of surface deposition due to depressurization accident for Cs-137 is 1.51E+03 Bq/m2 and due to water ingress accident for I-131 is 1.5E+01 Bq/m2. The impact of RDE radioactivity for hypothetical accidents on the KNS-II site shows lower than BAPETEN regulatory requirements.

Graphite aerosol release to the containment in a water ingress accident of high temperature gas-cooled reactor (HTGR)

Abstract As a design-basis accident, the water ingress due to the rupture of a single steam generator tube is a major threat for high-temperature gas-cooled reactors (HTGR). The rapid rise of primary pressure could reach the set pressure of safety valve and trigger the relief to the containment. For HTGR, the graphite dust particles, which usually carries fission product condensations, are also discharged to the containment during the pressure relief. The size distribution of the aerosol is crucial for the safety analysis of the containment in this accident scenario. Particularly, the high velocity flow in the safety valve may greatly change the size distribution of the graphite aerosol, due to the possible impact and fragmentation. In the present work we study graphite particle impact and fragment during pressure relief in the water ingress accident of HTGR by combining multiphase computational fluid dynamics (CFD) and finite element method (FEM). An Eulerian-Lagrangian scheme is employed to calculate the trajectory and instantaneous impact velocity of different sized particles. Then FEM with Hertz contact model is used to evaluate the outcome of the impact. CFD results show that the impact rate increases from 22% to nearly 100% as particle size increases from 0.1 μm to 1.9 μm. The critical fragment velocity is larger for smaller particle because the compressive strength limit increases due to Hall-Petch effect. By combining CFD and FEM simulation, it is shown that the possibility of fragmentation reaches 80% for particles larger than 1.5 μm. Our work suggests that this fragmentation effect needs to be considered for the initial size distribution of graphite aerosol that is discharged to the containment.

Analysis of helium purification system capability during water ingress accident in RDE

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

Abstract Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1 200 K. Three options of UO2, PuO2, and ThO2/UO2 fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO2-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO2/UO2-fueled pebble-bed HTR showed a similar trend to that of UO2, but did not show reactivity peak caused by water ingress. In contrast with UO2- and ThO2-fueled pebble-bed HTR, although the reactivity of PuO2-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0.10 cm thick SiC coating, it can be concluded that the effect of SiC coating on core reactivity in water ingress accident is more dominant for the pebble-bed HTR fuelled with thorium than those with uranium and plutonium fuels.

Stability of SiC and SiO<sub>2</sub> in Helium with Little Water Vapor

The stability of SiC and SiO2 in helium with little water vapor is studied in this paper in order to confirm whether they can service as anti-oxidation layers on the surface of fuel elements of high temperature gas-cooled reactor (HTR) under water-ingress accident. Through thermodynamic analysis, it is found that the influence factor controlling the critical temperature of passive oxidation for SiC in He-H2O is the partial pressure of H2O; the critical temperature of passive oxidation for SiC increases with the partial pressure of H2O. The influence factors controlling the critical temperature of decomposition for SiO2 are the partial pressure of H2O and the amount of H2O in gas for 1 mol SiO2, the critical temperature of decomposition for SiO2 increases with the partial pressure of H2O while it decreases with the amount of H2O in gas for 1 mol SiO2. SiO2/SiC composite coating is more suitable than SiC coating in He-H2O with little H2O.

Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of radioactive isotopes releasing to the secondary circuit during the blow-down transient will be prevented.

Recent Progress in Solid Breeder Blanket Development at JAEA

Japan Atomic Energy Agency (JAEA) plays a role of the principal institute in Japan for the design and the development of a solid breeder (WCSB) blanket and a helium cooled solid breeder (HCSB) blanket, in the ITER Test Blanket Modules (TBM) programt. The WCSB and HCSB modules consist of reduced activation ferritic/martensitic steel, F82H, as the structural material, Li2TiO3 as the tritium breeder material, beryllium or Be-Ti alloy as the neutron multiplier. One of the R&Ds for the WCSB TBM, the mockup of the first wall with embedded cooling channels was fabricated by applying HIP technique. Pebbles of Be12Ti, which is a candidate material for the advanced neutron multiplier, were produced by a small-scale rotating electrode method. Mechanical and chemical properties and irradiation effects have been studied for Be12T pebbles. Both oxidation and steam interaction were about 1/1000 as small as those of beryllium metal, which indicates a possibility to reduce a risk of a water or air ingress accident. The test schedule of TBMs is discussed according to the ITER operation phases.

Thermal hydraulic transient analysis of the HTR-10

The HTR-10 is the first high temperature gas-cooled test module reactor built in China. In the accident analysis, typical design basis accident and beyond design basis accident (BDBA), including the reactivity accident, the loss of external power ATWS, the air ingress accident and water ingress accident into the primary system are selected and detailed analyzed. The results show that the HTR-10 has inherent safety properties. The maximum fuel element temperature will not exceed the limit value 1230 °C. The total amount of graphite corrosion maintains no more than 320 kg, the exposure ratio of first coated particles is less than 2.4% in the BDBA. The released radioactivity is limited at a low level and the ability of fuel particles to retain fission products is not corrupted. Even the consequence of the severest hypothetical accident has no safety danger to the reactor. The reactor can shut itself down via its negative temperature coefficient of reactivity in heat-up conditions.

Fission product release and its environment impact for normal reactor operations and for relevant accidents

The radioactive concentration in the primary loop and the radioactive release for both normal operations and accidents for the HTR-10 are calculated and presented in the paper. The coated-particle fuel is used in the HTR-10, which has good performance of retaining fission products. Therefore the radioactive concentration in the primary loop of the HTR-10 is very low, and the amount of radioactive release to the environment is also very small for both normal operation and accident conditions. The radiation doses to the public caused by radioactive release for both normal operations and accidents are given in the paper. The results show that the maximum individual effective dose to the public due to the release of airborne radioactivity during normal operations is only 1.4×10 −4 mSv a −1, which is much lower than the dose limit (1 mSv a −1) stipulated by Chinese National Standard GB8703-86. For depressurization accident and water ingress accident, the maximum individual whole-body doses to man are only 7.7×10 −2 and 2.0×10 −1 mSv, thyroid doses only 1.7×10 −1 and 1.1 mSv, respectively. They are much lower than the prescribed minimum of emergency intervention level (whole-body dose: 5 mSv, thyroid dose: 50 mSv) for sheltering measures stipulated by the Chinese Nuclear Safety Criterion HAD002/03. The conclusion is that the environmental impact is very small for normal operations and accidents for the HTR-10, and the requirements stipulated in the Chinese Nuclear Safety Criterions are satisfied perfectly.

Safety Analysis Results for Cryostat Ingress Accidents in ITER

Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

Increasing the power of the high temperature reactor module

To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoreticallybelow 1600 °C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%.