Depressurization Accident Research Articles

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.

Experiment Study on Rod Ejection Accident of Hydraulic Control Rod Driving System

The rod ejection experiment has been done in the 1:1 experiment loop of the hydraulic control rod driving system (HCRDS) of the 200 MW nuclear heating reactor (NHR-200). The result indicated that under different experiment conditions, the control rod was not ejected when the pressure vessel of the loop lost pressure. Based on the depressurization accident analysis of the NHR-200, performed at the Institute of Nuclear Energy Technology (INET), Tsinghua University, the depressurization rate of experiment is much higher than that of analysis. Thus, the HCRDS has inherent safety and no hidden trouble of rod ejection when the depressurization accident happens.

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR

It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows:1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS.2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS.3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels.The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Experiment Study on Rod Ejection Accident of Hydraulic Control Rod Driving System.

The rod ejection experiment has been done in the 1:1 experiment loop of the hydraulic control rod driving system (HCRDS) of the 200 MW nuclear heating reactor (NHR-200). The result indicated that under different experiment conditions, the control rod was not ejected when the pressure vessel of the loop lost pressure. Based on the depressurization accident analysis of the NHR-200, performed at the Institute of Nuclear Energy Technology (INET), Tsinghua University, the depressurization rate of experiment is much higher than that of analysis. Thus, the HCRDS has inherent safety and no hidden trouble of rod ejection when the depressurization accident happens.

Advanced Vessel Cooling System Concept for High-Temperature Gas-Cooled Reactors

The high-temperature gas-cooled reactor (HTGR) program will be attractive to a broad range of owner / operators and meet public acceptance if the future HTGRs would be completely free from accidents, which cause a significant release of radioactivity into the environment. An advanced vessel cooling system concept, in which there is no heat loss in normal operation and the decay heat is removed by the natural circulation of air in an accident, is proposed for the High-Temperature Engineering Test Reactor to meet this requirement. The depressurization accident, one of the severest accidents of the HTGR, is selected and the analysis shows no significant core heatup. Applicability to the future HTGR is also investigated.

Temperature equalization phenomenon in a high temperature reactor core with central column

Abstract The usual design of a High Temperature Reactor in Germany refers to a pebble bed core of cylindrical shape, like AVR, THTR and HTR-Modul. Since also other geometries are possible, concept considerations for a HTR with non-cylindrical core have already been made, indicating some advantage in safety due to a limitation of the maximum core temperature in case of a depressurization accident of the primary loop. Using the concept of the AHTR-500 for process heat applications the effectiveness and influence of the central column was evaluated. Results for heat flows and stored heat in core and central column have been calculated for the case of depressurization by using the computer codes THERMIX/DIREKT/TPlot.

Experimental Study of Dust Behavior during Depressurization.

A small amount of graphite dust is present in the primary cooling system of a graphite, moderated gas-cooled reactor during normal operation and condensable fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. In order to obtain liftoff fractions (fractional amount removed) of the graphite dust during the depressurization, an investigation was performed by the blow down test. In the test, graphite powder attached on the sample was blown down under the high speed gas flow to simulate flow condition during the depressurization accident. According to the shear ratio concept, the liftoff fraction can be described by shear ratio, which is defined as the ratio of wall shear stress in the blow down condition to the normal condition. From the test result, it is concluded that the shear ratio concept is not completely en...

Safety Analysis of Depressurization Accidents in a Medium-Sized High-Temperature Gas Reactor

In this paper the potential release of fission products during a beyond-design accident in a medium-sized high-temperature gas reactor (the HTR-500) is investigated. The DSNP modular simulation code is used to simulate a depressurization accident as well as the failure of the forced circulation of the decay heat removal systems to actuate. For such an extreme accident, the calculated maximum localized fuel temperature reaches 3040{degrees} C 43 h after the beginning of the accident. During the heatup, 3.4% of the {sup 137}Cs inventory is found to be released from the fuel elements to the primary circuit, and 4.6 {times} 10{sup {minus}2}% is estimated to be released into the environment. The carbon monoxide and helium releases from the graphite matrix prove to be an important factor in sweeping the fission products from the primary circuit. The comparative consequence analysis indicates a much lower risk than in the analogous light water reactor severe accident. A design-base depressurization accident is also investigated at the beginning of the study and involves the operation of one out of the two redundant decay heat removal systems.

Depressorization accident analysis for the FTTR by the TAC-NC

The two-dimensional thermal analysis code TAC-NC is modified from the analytical code TAC-2D in order to calculate temperature transients in the case of loss of forced cooling accidents of the HTTR (High Temperature engineering Test Reactor) such as a depressurization accident. The TAC-NC code includes a special function to calculate heat transfer by natural convection between hotter and colder regions in the pressure vessel during the depressurization accident. Verification of the TAC-NC code was carried out by the comparison of the analytical results with the experimental ones of an air ingress test. Analytical results of simulated core temperature were in good agreement with experimental results. Temperature transients during a depressurization accident were evaluated by the TAC-NC code for the HTTR. The maximum fuel temperature decreases rapidly after the reactor scram and increases slightly after that due to decay heat. The maximum fuel temperature, however, remains below the initial maximum fuel temperature since most of the core decay heat is absorbed in the large thermal capacity of graphite in the core and reflector.

Studies on the concepts of HTGR plants viable in Japan Part II: Discussion of the results

Two topics are reviewed as the major results of the studies reported here: assessing the mechanisms of the inherent safety characteristics of the Modular HTGRs. Comparison of specific features between HTR-Module and MHTGR and the analyses of the heat transport capability at the depressurization accident are included. The reference plant is examined for the viability assessment studies. In this paper, some of the results of our scenario for the design of the PCRV type HTGRs are introduced.

Experimental and theoretical investigations of creep buckling on NiCr 22 Co 12 Mo tubes at 950°C

The postulated secondary-side depressurization accident in an HTR plant raises the question of whether creep buckling can occur under external pressure in the tubes of the heat exchangers whose cross sections may be slightly eccentric due to manufactural tolerances. Experimental and theoretical investigations were carried out on the problem of creep buckling. Within the scope of these experiments, tubes made of NiCr 22 Co 12 Mo with different adjusted initial ovalities were exposed to pressures of 35–45 bar at 950°C. The buckling times were measured. Theoretical analyses were carried out using a model conceived by Hoff as well as the Finite Element Method, allowing for the scattering of relevant test parameters. Both methods of computation are suitable for investigating the phenomenon of creep buckling. It was found that the calculated buckling times are affected by great uncertainties taking into account the scatter of influential test parameters. In connection with the accidental stressing in an HTR plant it was confirmed that creep buckling is not a problem of high safety-related relevance.

Fuel management effects on inherent safety of modular high temperature reactor.

Analysis was performed on the effects of fuel loading schemes and fuel materials on the inherent (passive) safety characteristic of the modular pebble-bed type high-temperature reactor against depressurization accident involving loss of He forced circulation Two extreme fuel loading schemes, the infinite-velocity multipass and Once-Through-Then-Out (OTTO), were evaluated for both U and Th fuel cycles. The results of the analysis show that the maximum core temperatures attained following the accident were much lower for the infinite-velocity multipass scheme than for the OTTO scheme. For both schemes, the Th cycle showed slightly higher maximum peak temperatures, compared to U cycle.

Radiological consequences of a depressurization accident combined with water ingress in an HTR module-200

The principles of the reaction between Cs and steam or water with respect to the primary circuit of an HTR and results of corresponding experiments are presented. Based on this for the most important pressure loss accidents combined with water ingress of an HTR-200 the radiological consequences for the environment are investigated in detail. The results are presented and discussed.

GHR 10 MW: The technical concept of the gas cooled heating reactor

The GHR is a simple, safe and economic heating reactor that is designed on the basis of the proven technology of the Gas-Cooled High-Temperature Reactor. In the HRB/BBC-design of the GHR the basic proven HTR components are: • - the spherical HTR fuel elements containing coated particles for an extremely favourable fission product retention. • - the primary circuit with helium coolant and ceramic core structures integrated into a prestressed concrete reactor vessel using the liner as the main heat exchanger. The main benefits are derived directly from the favourable technical and safety related HTR-characteristics. These are • - low radiation exposure, • - high inherent safety, even in the event of hypothetical accidents. Even in the hypothetical case of a depressurization accident with total loss of active cooling the core temperatures will remain below 800°C. The corresponding equivalent whole body dose will be lower than 10 mrem. Due to these favourable safety characteristics, GHR plants can be sited near urban centers without any risk to the population.

Calculation of the depletion of a radioactive plume in the atmosphere and subsequent exposure due to deposited activity

An efficient method has been developed for calculating the depletion, by dry deposition and washout to the ground, from a plume of fission products in the atmosphere, and for determining the isotopic composition of resulting gamma and beta activities in the ground deposit at any selected point downwind. By integration over the pattern of material on the ground, gamma and beta exposure rates may be derived as a function of the ageing time of the deposit. Elements in the plume are assigned to one of up to four ‘groups’; elements within each group being assigned a particular deposition velocity and washout coefficient. Calculations have been performed, for a notional MAGNOX depressurization accident, which illustrate the sensitivity of ground gamma spectra and exposure rate to the use of isotope-dependent deposition velocities. Circumstances under which it may be necessary to perform a full two-dimensional integration over the pattern of material on the ground are discussed.

Transient Analyses of a 1000-MW Gas-Cooled Fast Reactor

To calculate the depressurization and flow-coast-down accidents in a 1000-MW gas-cooled fast reactor (GCFR) with a secondary steam cycle, the PHAETON2 computer code is used, the emphasis being placed on the solution of one-dimensional unstationary helium flows. The fluid dynamics equations are solved one by one by a combination of implicit and explicit methods, taking into account most of the terms of the original equations. In the case of the accidents considered, the shutdown system is always activated, and inherent actions only of the GCFR are allowed. The results show a necessity of backup pressures above 150 kPa for the depressurization accidents and a minimum circulator frequency of 5 Hz for the flow-coastdown accidents.

Analysis of the Venting System of the Gas-Cooled Fast Breeder Reactor Fuel Element for Various Operating Conditions

The fuel element concept of the gas-cooled fast breeder reactor (GCFR) is based on vented fuel pins to equalize pressure differences between the fission gas inside the fuel pin and the coolant. The fission products escaping from the fuel, mainly noble gases, are collected and swept separately from the primary coolant by a helium stream into a purification plant. Calculations were performed to estimate the activity release during normal operation, transient, and accident conditions for a 1000-MW(e) GCFR designed by Kraftwerk Union. The results show that during normal operation, only 0.8% of the total equilibrium noble gas activity in the core will be released into the purification plant. The most severe case for the activity release is a depressurization accident followed by the release of the whole fission gas inventory in the interstitial gas volume of the fuel pins of ∼5.3 × 107 Ci (2.0 EBq). To adsorb this amount of fission gases in the low-temperature charcoal beds of the purification plant, a temporary refrigeration load of ∼173 kW is necessary. Using a purification plant with a refrigeration capacity of ∼50 kW and an equivalent storage of liquid nitrogen for auxiliary purposes, no significant extrapolation from the designed high-temperature gas-cooled reactor purification plants is necessary.

Reliability Analysis of the Decay Heat Removal System of a 1000-MW(e) Gas-Cooled Fast Breeder Reactor

A reliability analysis for the decay heat removal system of a 1000-MW(e) gas-cooled fast breeder reactor (GCFBR) has been carried out. It relates only to the GCFBR specific systems required after the maximum depressurization accidentnormal shutdown. Non-GCFBR specific systems such as the electrical supply system and the containment back-pressure system were not included in the analysis. For these analyses, the method of fault trees was used to describe the different failure combinations of the system. The calculations were performed with the system analysis program SAP-1, which combines analytical and simulation methods to find the failure probability and the unavailability of the system. The results of the analysis show that the unavailability governs the reliability of the system; a computed value of 3 × 10−4 was obtained for the unavailability of the system after a maximum depressurization accident, and a value of 4 × 10−8 was obtained for the unavailability of the system after normal shutdown.

An analysis of GCFR depressurization transients without scram

Four transients imposed on a gas cooled fast breeder reactor (GCFBR or GCFR) plant are analyzed. The transients are variations of a design basis depressurization accident (DBDA) [1], in which a rupture is postulated in the inlet plenum, the reactor is scramed and the circulator output is drastically reduced. Variations considered are, (1) in all cases the reactor is not scrammed; and (2) the rupture size is varied from 0.6 ft 2 to zero. In the limit of zero (no rupture) the transient imposed on the system is due to the behavior of the circulator and steam generator during a DBDA.

Effects of Helium Mixing and Heat Transfer on Containment Design Pressure in a High-Temperature Gas-Cooled Reactor

Containment design pressure for a high-temperature gas-cooled reactor is determined by its response to a design basis depressurization accident. The effects of heat transfer to internal structures and of helium mixing significantly affect the response. In the mathematical model discussed, the containment is divided into two regions; a lower region that contains only air, and an upper region that contains all the helium and whatever air is assumed to mix. Heat sinks are distributed vertically. At each instant, a given heat sink is calculated to be in either the unmixed region or the mixed region. In this way, both the mixing fraction and the heat transfer data can be changed. The peak pressure can be reduced by (a) placing heat sinks higher in the containment, (b) increasing the mixing fraction, and (c) accounting for heat transfer as the helium rises through the lower region.