Hypothetical Accident Research Articles

Emergency preparedness of Research Center for Radiation Medicine and its hospital to admit and treat the patients with signs of acute radiation sickness

After the Chernobyl accident, the Research Center for Radiation Medicine (RCRM) was established in Kiev (Ukraine). Its main task was to maintain a high level of emergency preparedness and be ready to examine and treat patients who suffer as a result of hypothetical radiation accident. Based on the previous experience, this institution's specialists worked out new diagnostic criteria and drug treatment schemata for acute radiation sickness, created a database on 75 patients with this diagnosis and improved educational programmes for medical students and physicians working in the field of radiation medicine. RCRM collaborates fruitfully with western partners through the joint research projects and connects with the World Health Organization's Radiation Emergency Medical Preparedness and Assistance Network centre. Collaboration with Kiev Center for Bone Marrow Transplantation allows RCRM to use aseptic wards having highly filtered air for the treatment of most severely irradiated patients.

Study of heat and mass transfers in a spray for containment application: Analysis of the influence of the spray mass flow rate

During the course of a hypothetical accident in a nuclear power plant, spray might be activated in order to reduce static pressure in the containment. The IRSN has developed the TOSQAN experiment to give a better understanding of the heat and mass transfers between a spray and the surrounding confined gas. This article analyses the influence of the injected spray mass flow rate on the thermal-hydraulics of spray tests. The three tests compared in this article, to carry out this analysis, have the same boundary and initial conditions, except the spray mass flow rate ranges from 10 to 53 g s −1. First, the scenario of these three tests used for this analysis and their results are presented. Then, our analysis is focused on the inter-comparison of the thermal-hydraulic behavior induced by spray mass flow rate variations. This inter-comparison is divided into two parts: a global and a local one.

Radiation consequences of hypothetical accidents associated with transportation of spent nuclear fuel of nuclear submarines aboard floating technical base

Reload of spent nuclear fuel of nuclear submarines and its subsequent transportation for reprocessing may be carried out by the vessel – floating technical base (FTB). Up to 4–5 reactor cores (∼ 10 16 Bq) can be taken aboard, transportation can occur for long distance. In case of the accidents with infringement of protective barriers, FTB may be a significant source of radiation hazard. Two hypothetical accidents associated with explosion and fire aboard and sinking the FTB are analyzed. In the first case a source term for the release and aerial radionuclide spreading out have been estimated, in second case radiation consequences have been estimated for water way taking into account remaining protective barriers. Main task of the assessments was radiation impact on population and environment at long distances from the source.

Radioecological consequences of a potential accident during transport of spent nuclear fuel along an Arctic coastline

This article presents results pertaining to a risk assessment of the potential consequences of a hypothetical accident occurring during the transportation by ship of spent nuclear fuel (SNF) along an Arctic coastline. The findings are based on modelling of potential releases of radionuclides, radionuclide transport and uptake in the marine environment. Modelling work has been done using a revised box model developed at the Norwegian Radiation Protection Authority. Evaluation of the radioecological consequences of a potential accident in the southern part of the Norwegian Current has been made on the basis of calculated collective dose to man, individual doses for the critical group, concentrations of radionuclides in seafood and doses to marine organisms. The results of the calculations indicate a large variability in the investigated parameters above mentioned. On the basis of the calculated parameters the maximum total activity (“accepted accident activity”) in the ship, when the parameters that describe the consequences after the examined potential accident are still in agreement with the recommendations and criterions for protection of the human population and the environment, has been evaluated.

Impact target characterisation of BAM drop test facility

BAM safety related research of containers for radioactive material focuses on advanced mechanical safety assessment methods for verification of the structural integrity and leaktightness under normal and hypothetical accident conditions during transport and storage. An essentially unyielding target with a rigid surface is required for impact tests performed for package approval according to IAEA regulations [1]. In addition to specification of a target, e. g. with a combined mass more than 10 times that of the specimen for drop tests, unyielding target characteristics have been investigated with various package designs and different impact tests. The unyielding target of the BAM drop test facility, a reinforced concrete block together with an embedded and anchored mild steel plate, provides relatively large mass and stiffness with respect to the packages being tested. For monitoring reasons accelerometers and strain gauges are embedded in the concrete block of the foundation at several positions. Additionally, dynamic impact responses like vibrations and rigid body motion can be measured by seismic accelerometers. The mechanical characterization of the target’s rigidity is based on experimental results from various drop tests. Test containers with weights of 181,000 kg, 127,000 kg and 8,010 kg hit the target with velocities up to 13.5 m/s in the horizontal and vertical drop positions. The rigidity of the impact target can be demonstrated with experimental results confirmed by analytical approaches. Some conclusions can be drawn about experimental testing as well as analytical calculations in order to compare impact effects.

Potential Threats from a Likely Nuclear Power Plant Accident: a Climatological Trajectory Analysis and Tracer Study

The legacy of Chernobyl is not the only nuclear accident likely to confront Turkish territory, which is not far from other insecure power plants, especially the Metsamor. The main purpose of this study was to examine the possible impacts to Turkish territory of a hypothetical accident at the Metsamor Nuclear Plant. The research was performed based on two different methodologies: First, a 10-day trajectory analysis was carried out a set of long-term (30 years) meteorological data; second, a tracer study was performed using the MM5T online model for the selected episode. Trajectory and tracer studies showed that an accident at the Metsamor Nuclear Power Plant would influence all of Turkey. Furthermore, vulnerable regions in Turkey after the Chernobyl disaster were demonstrated as a new and first attempt in this study.

Analysis of heat and mass transfers in two-phase flow by coupling optical diagnostic techniques

During the course of a hypothetical accident in a nuclear power plant, spraying might be actuated to reduce static pressure in the containment. To acquire a better understanding of the heat and mass transfers between a spray and the surrounding confined gas, non-intrusive optical measurements have to be carried out simultaneously on both phases. The coupling of global rainbow refractometry with out-of-focus imaging and spontaneous Raman scattering spectroscopy allows us to calculate the local Spalding parameter BM, which is useful in describing heat transfer associated with two-phase flow.

Time Evolution Simulation for a Realistic In-containment Source Term During Hypothetical Accidents

The time evolution of an iodine (as a key and representative volatile fission product species) transport and removal is simulated to evaluate a realistic in-containment source term instead of TID-14844 under a conservative hypothetical scenario in PHWRs. Fission products in vapor and aerosol forms can be deposited onto every heat sink in a compartment and be dissolved in a water pool by different deposition mechanisms. All these reactions governing an iodine development and evolution in a containment under severe accident conditions are simulated by using the latest ISAAC 2.x version which includes modified iodine release/transport/deposition models.

Development of Health Effect Assessment Software Using MACCS2 Code

The extended regulatory interests in severe accidents management and enhanced safety regulatory requirements raise a need of more accurate analysis of the effect to the public health by users with diverse disciplines. This facilitates this work to develop web-based radiation health effect assessment software, RASUM, by using the MACCS2 code and HTML language to provide diverse users (regulators, operators, and public) with easy understanding, modeling, calculating, analyzing, documenting and reporting of the radiation health effect under hypothetical severe accidents. The engine of the web-based RASUM uses the MACCS2 as a base code developed by NRC and is composed of five modules such as development module, PSA training module, output module, input data module (source term, population distribution, meteorological data, etc.), and MACCS2 run module. For verification and demonstration of the RASUM, the offsite consequence analysis using the RASUM frame is performed for such as early fatality risk organ does, and whole body does for two selected scenarios. Moreover, CCDF results from the RASUM for KSNP and CANDU type reactors are presented and compared.

Recent assessments in USA of spent fuel packages exposed to severe thermal environments different from regulatory standards

AbstractThe regulatory driven design of radioactive material transportation packages leads package vendors to perform analyses that demonstrate the ability of packages to meet the regulatory requirements. For risk assessment and communication, the analysis of package response to thermal environments that are more severe than those described in the regulations is required. In general, experimental and analytical assessments of casks exposed to thermal insults other than the regulatory environment are performed in the USA by the Department of Energy national laboratories. This paper provides a brief summary of some recent thermal analyses of spent fuel transportation packages exposed to thermal environments different from regulatory standards. The analyses were performed by Sandia National Laboratories under several different projects for multiple customers. These analyses examined the response of spent fuel packages exposed to severe thermal environments different from the regulatory hypothetical accident ...

LBE–water interaction in sub-critical reactors: First experimental and modelling results

This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

Implementation of the results of the uncertainty analysis of the COSYMA package for zoning purposes

The EC and US probabilistic accident consequence codes, COSYMA and MACCS, respectively, estimate the risks and other endpoints associated with hypothetical accidents at nuclear installations. A joint project for an uncertainty analysis of these two codes used a rigorous expert judgement procedure with multiple experts. For all assessments, eight panels of six to ten experts were used. The uncertainty analysis of COSYMA evaluated the uncertainty on 191 endpoints for three different situations that are of relevance to regulatory processes in different European countries. The uncertainties were evaluated for three source terms. In this way the study encompassed a wide range of source terms and other situations that have been considered in safety assessments of nuclear power plants. This paper will mainly deal with two aspects: (1) relating predictions using default parameter values and providing a basis for determining where the 95th envelope in the uncertainty analysis lies and (2) how to estimate uncertainties in a level-3 Probabilistic Safety Assessment (PSA) by applying the results of the uncertainty study for other situations. For individual risk, an example will be shown for zoning around nuclear power plants.

Analysis of Noncondensable Gas Recirculation Flow in Steam Generator U-Tubes during Reflux Condensation Using RELAP5

Under certain hypothetical accident circumstances in a pressurized water reactor, reflux condensation in steam generator U-tubes may be an effective heat removal mechanism. In a previous study, an empirical correlation of condensation heat transfer coefficients for reflux condensation was derived and incorporated into the transient analysis code RELAP5. In this study, different flow patterns in U-tubes observed in the reflux condensation test 7.2c with nitrogen injection in the Bethsy facility were calculated using a parallelchannel model for steady-state calculations and the RELAP5 code for transient calculations. In the region of low steam flow rates, the pressure in the outlet plenum becomes higher than that in the inlet plenum due to the density head of gases, and nitrogen reverse flow occurs, which was successfully calculated using the RELAP5 code with modified equations for the interfacial and wall friction forces. There are many solutions with different numbers of active U-tubes with steam condensation and the predicted number of active U-tubes agrees well with the measured values based on the assumption that the flow might be most stable in the case with the maximum nitrogen recirculation flow rate.

A pressure-tube Advanced Burner Test Reactor concept

A new steady-state fast neutron test reactor has been conceptually designed. This paper presents a new concept for a CANDU-based fast test reactor that is horizontal in orientation, with individual pressure/Calandria tubes (PT/CT) running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely reconfigure the core to meet different users’ requirements. Full core neutronic analysis showed a small hexagonal reactor, LBE-cooled, trans-uranics (TRU)-67Zr fuel with HT-9 cladding and structures, with a core power of 100 MW th produced a fast flux (>0.1 MeV) of 1.4 × 10 15 n/cm 2 s averaged over the whole length of six irradiation channels with a total testing volume of more than 77 l. Loading of TRU from legacy UO 2 spent fuel allowed core continuous operation for 180 effective full power days with a net fissile-Pu burning rate of ∼6.4%. Since high neutron fluence impact on PT/CT might be an issue of concern for this design, oxide dispersion strengthened (ODS) ferritic steel was used as PT/CT material without any impact on core neutronic behavior. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. A preliminary thermal-hydraulic analysis of LBE produced acceptable pumping power and clad temperature. Voiding the core from its LBE coolant resulted in a positive reactivity insertion which is typical of most fast reactors. Hypothetical accidents of draining the Calandria and/or the external LBE reflector tank of the LBE core resulted in negative reactivity insertion which shut the reactor down.

Prediction of TRISO coated particle performances for a one-pass deep burn

In the present studies, TRISO coated particle performances have been investigated for incinerating plutonium and minor actinides by the Gas Turbine-Modular Helium Reactor, whose fresh fuel is fabricated after the uranium extraction (UREX) process applied to Light Water Reactors irradiated fuel. The analyses divide into two parts: in the first part, the latest design of the reactor core proposed by General Atomics, which takes advantage of four fuel rings, has been modeled in deep details by the Monte Carlo MCNP code and a burnup process has been simulated by the MCB code. In the second part, the TRISO coated particle performances have been investigated by the PANAMA code with the goal of verifying the design constraints proposed by General Atomics. During burnup, the refueling and shuffling schedule followed the one-pass deep burn concept, where the fuel is utilized, since fabrication for the Gas Turbine-Modular Helium Reactor, without any reprocessing until the final disposal into the geological repository. During the reactor operation, the fast fluence on all TRISO particles layers has been evaluated and the production of the key fission products monitored. During an hypothetical reactor accident scenario, the TRISO particle failure fraction has been estimated.

HI-STAR 100 spent fuel transport cask analytical evaluation for drop events

The US Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (CFR), Title 10, Part 71. Both the International Atomic Energy Agency (IAEA) regulations for transporting radioactive materials (Ref. 2, paragraph 727) and 10 CFR 71·73 require casks to be evaluated for hypothetical accident conditions, which includes a 9 m (30 feet) drop impact event on a flat, essentially unyielding, horizontal surface, in the most damaging orientation. The present paper examines the behaviour of one of the NRC certified transportation casks, the HI-STAR 100, for drop impact events. The specific area examined is the behaviour of the bolted connections in the 'overpack' top flange and the closure plate, which are significantly loaded during the hypothetical drop impact event. The term 'overpack' refers to the cask that receives and contains a sealed multipurpose canister (MPC) containing spent nuclear fuel. The analytical work to evaluate the NRC certified HI-STAR 100 spent fuel transport cask for a 9 m (30 feet) drop impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS and LS-DYNA finite element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9 m (30 feet) free drop hypothetical accident conditions, as specified in 10 CFR 71·73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1 m (40 inch) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71·73, was not included in the current work. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the closure plate surfaces may occur, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.

TN-68 spent fuel transport cask analytical evaluation for drop events

AbstractThe US Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71. Both the International Atomic Energy Agency regulations for transporting radioactive materials, and 10 CFR 71·73 require casks to be evaluated for hypothetical accident conditions, which includes a 9 m (30 ft) drop impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behaviour of one of the NRC certified transportation casks, the TN-68, for drop impact events. The specific area examined is the behaviour of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop impact event. Analytical work to evaluate the NRC certified TN-68 spent fuel transport cask for a 9 m (30 ft) drop impact event on a flat, unyielding, horizontal surface, was performed using the ANSY...

Reflux Condensation Heat Transfer of Steam-Air Mixture under Turbulent Flow Conditions in a Vertical Tube

Under certain hypothetical accident circumstances in a pressurized water reactor, reflux condensation in the steam generator U-tubes may be an effective heat removal mechanism. Non-condensable gas (air) may be present in the reactor coolant system and is well known to inhibit steam from condensing. In a previous study, local condensation heat transfer coefficients for reflux condensation with air were measured, and an empirical correlation intended for a laminar flow of steam-air mixtures was developed, which was a function of the partial pressure ratio, (P steam=P air). In this study, heat transfer coefficients measured under turbulent flows were used to modify the empirical correlation for improved prediction accuracy. The modified empirical correlation considers convection and is proportional to (P steam/P air)0.75 Re g 0.8 where Re g is the steam-air mixture Reynolds number. The correlation was verified against the data of Moon et al. and then incorporated into the transient analysis code RELAP5/SCDAPSIM/MOD3.2. The temperature distributions of the steam-air mixture calculated by the code for six reflux condensation tests are shown to agree well with the measured results.

The GIS-based SafeAirView software for the concentration assessment of radioactive pollutants after an accidental release

The European Commission Joint Research Centre (JRC) in Ispra (Italy) has long been running nuclear installations for research purposes. The Nuclear Decommissioning and Facilities Management Unit (NDFM) is responsible for the surveillance of radioactivity levels in nuclear emergency conditions. The NDFM Unit has commissioned the implementation of a specifically developed decision support system, which can be used for quick emergency evaluation in the case of hypothetical accident and for emergency exercises. The requisites were to be a user-friendly software, able to quickly calculate and display values of air and ground radioactive contamination in the complex area around JRC, following an accidental release of radioactive substances from a JRC nuclear research installation. The developed software, named "SafeAirView", is an advanced implementation of GIS technology applied to an existing MS-DOS mode dispersion model, SAFE_AIR (Simulation of Air pollution From Emissions_Above Inhomogeneous Regions). SAFE_AIR is a numerical model which simulates transport, diffusion, and deposition of airborne pollutants emitted in the low atmosphere above complex orography at both local and regional scale, under non-stationary and inhomogeneous emission and meteorological conditions. SafeAirView makes use of user-friendly MS-Windows type interface which drives the dispersion model by a sequential and continuous input-output process, allowing a real time simulation. The GIS environment allows a direct interaction with the territory elements in which the simulation takes place, using data for the JRC Ispra region represented in geo-referenced cartography. Furthermore it offers the possibility to relate concentrations with population distribution and other geo-referenced maps, in a geographic view. Output concentration and deposition patterns can be plotted and/or exported. In spite of the selected specific databases, the SafeAirView software architecture is a general structure, therefore the decision support system could be easily modified to be applied in a region different from the JRC one. Beside the description of SafeAirView, the present paper presents a statistical evaluation of the software, which has considered three well known tracer experiments: Copenhagen, Indianapolis and Kincaid. The data sets related to these experiments are all included in the so called Model Validation Kit.

Thermal testing of type B packages in furnaces per ASTM standard practice E 2230

During 2003, staff of the Transportation Technologies Group of Oak Ridge National Laboratory (ORNL) conducted hypothetical accident conditions (HAC) thermal tests, as specified in 10 CFR 71·73 (c)(4), on two separate packages – the DPP-2, MD-1, ES-2100 and the ES-3100. The basis for each of these tests was the newly published ASTM standard practice E 2230 entitled 'Standard practice for thermal qualification of type B packages for radioactive material'. These two series of tests represent the first known use of this ASTM standard which was authored by a panel of experts in the fields of thermal testing and thermal analysis, including participation by the US Nuclear Regulatory Commission. This new ASTM standard was initially adopted in 2002 and first published in 2003. The purpose of this paper is to describe the results and key aspects of conducting type B thermal tests using furnaces. These tests were performed at commercial heat treat facilities using large gas fired furnaces. Before actual testing, each of the furnaces, as well as all of the tested units, were fully instrumented to provide data to be used as the basis for showing compliance with E 2230 as well as 10 CFR 71. Additionally, a data acquisition system, capable of continuously monitoring up to 72 channels of data, was developed at ORNL specifically for use during these tests. The paper highlights the various techniques employed to ensure compliance with both 10 CFR 71 and ASTM E 2230. These include the technical basis for testing activities, logistics of conducting off site testing, the preheat process each of the packages was exposed to, detailed descriptions of instrumentation used and procedures employed during testing, as well as lessons learned from the testing activities.