Nuclear Safety Institute Research Articles

An experimental study of the thermal-hydraulics of a spent nuclear fuel pool under loss-of-cooling accident conditions. Part I: The MIDI facility and its main experimental results

Motivated by the 2011 accident at the Fukushima Daiichi nuclear power station, the French Institute for Radiation Protection and Nuclear Safety (IRSN) launched in 2014 the DENOPI project. This research program, supported by the French Government and carried out in collaboration with national and international partners, dealt with the open issues of spent fuel pool accidents. As part of this project, a test facility named MIDI was designed and built at the IRSN’s research unit of Cadarache, France. The MIDI facility is an experimental tool which aims at reproducing the thermal–hydraulic phenomena that may occur at an integral scale in a spent fuel pool undergoing a loss-of-cooling accident, before the stored fuels get uncovered. This paper first describes the MIDI facility and its instrumentation and the test matrix achieved to date. An overview of its main results is then provided. In the framework of the DENOPI project, nine tests were achieved and allowed investigating the phenomenology of this type of accident. The highlighted phenomenology is characterized by three competing liquid vaporization modes: the evaporation at the pool free surface, the nucleate boiling within the fuel bundles and the gravity-driven flashing of superheated water on top of the storage racks. The existence of the latter vaporization mode, uncertain at the launch of the DENOPI project, is now confirmed in the configuration of a spent fuel pool.

Comparative Study of Radiation Mapping Technologies for Nuclear Disaster Assessment

The distribution of the ambient dose equivalent rate (i.e., air dose rate) after a nuclear disaster is crucial for zoning contaminated areas to facilitate authorities’ effective decision making. Several countries are considering a gradual characterization strategy where airborne measurement is performed first followed by ground measurement (i.e., via manborne or carborne surveys). Nonetheless, potential differences might emerge in country-specific air dose rate assessment methods. Explaining these discrepancies can improve and converge existing methodologies. The Japan Atomic Energy Agency (JAEA) and the French Institute for Radiological Protection and Nuclear Safety (IRSN), which are organizations involved in post-nuclear accident crisis management, jointly performed air dose rate measurements in 2019 at contaminated sites around the Fukushima Daiichi Nuclear Power Station. The similarities and differences between the two organizations’ methods and results were quantitatively assessed by comparing the average air dose rates obtained within a grid created with a geographic information system, and the reasons for the differences between the organizations’ results were investigated. The air dose rates obtained by the manborne measurements varied depending on the calibration method. Comparing the air dose rate assessment methods and mapping techniques used in different countries will contribute to developing international guidelines for recommending the best method for determining air dose rates.

The European intercomparison of in-vivo monitoring laboratories: the EIVIC-2020 project

The EIVIC project was launched in 2020, and the main goal was the organisation of a European intercomparison of in-vivo monitoring laboratories dealing with direct measurements of gamma-emitting radionuclides incorporated into the body of exposed workers. This project was organised jointly by members of EURADOS Working Group 7 on internal dosimetry (WG7), the Federal Office for Radiation Protection (BfS, Germany) and the Radioprotection and Nuclear Safety Institute (IRSN, France). The objective was to assess the implementation of individual-monitoring requirements in EU Member States on the basis of in-vivo measurements and to gain insight into the performance of in-vivo measurements using whole-body counters. In this context, a total of 41 in-vivo monitoring laboratories from 21 countries, together with JRC (EC) and IAEA participated. The results were submitted in terms of activity (Bq) of the radionuclides identified inside phantoms that were circulated to all participants. The measured data were compared with reference activity values to evaluate the corresponding bias according to the standards ISO 28218 and ISO 13528. In general, the results of the different exercises are good, and most facilities are in conformity with the criteria for the bias and z-scores in the ISO standards. Furthermore, information about technical and organisational characteristics of the participating laboratories was collected to test if they had a significant influence on the reported results.

Multi-material topology optimization of nuclear devices – principle and application

In this paper, we present a multi-material topology optimization procedure, capable of generating the best possible structure of a nuclear device from a list of design objectives and constraints. We propose an example of applying this new tool to calculate the structure of a neutron moderator, which will be used to generate an epithermal field for metrological applications at a facility of the Institute for Radiation Protection and Nuclear Safety (IRSN, France).

Strengths of ecosystem services concept for radiation protection.

The successful ecosystem services concept, defined as the benefits people obtain from ecosystems is still not really reflected in the current approaches for protecting public and environment against radiation promoted by the International Commission on Radiological Protection or other similar approaches. Yet some recent thoughts from international organizations lead us to believe that an eco-based approach could be more promoted in the coming years in environmental radiation protection field. The French Institute for Radiation Protection and Nuclear Safety has identified different fields of application of this concept into radiation protection, in line with its integrated approach of radiological risks management. As the ecosystem services approach makes it possible to highlight biophysical and socio-economic approaches of the impacts of ionizing radiation on ecosystems, it represents a subject of primary importance for future works conducted by IRSN. However, the operationality of the ecosystem services concept is the subject of many debates. In many situations, scientists have not yet fully understood how radioactive contamination could affect ecosystem services, and how to articulate with certainty cause and effect relationships between state of an ecosystem and provision of services. In addition, the concept is also accompanied by contradictory perceptions of the status of humans in ecosystems. To solve these knowledge gaps and uncertainties, it is necessary to acquire robust data on the impacts of radiation on ecosystems both under experimental and realistic conditions, and to integrate all potential consequences (direct and indirect, ecotoxicological but also economic and cultural).

3D structure tectonics model of Yenisei site of the Nizhnekansk Massif

In 2021, in the Yeniseisky region of the Nizhnekansk Massif in the Krasnoyarsk region, construction of an underground research laboratory (URL) was started to justify safety of high-level radioactive waste disposal. The URL investigations aimed to assess preservation of isolation properties of rock mass exposed to long-active rock pressure, tectonic stresses and heat flow within the whole period of effective radiobiological risk of radioactive waste. Based on the geological data analysis, the structure-and-tectonics model of the Yenisei site was developed, including tectonic faults, lithology, intrusives, as well as heavy fracturing and crushing zones. With no large-scale acquisition of geomechanical data on the Lower Kan Massif, the rock stability was estimated at outcrops on ground surface. The results proved the existing hypotheses of the dynamic impact exerted by the Muratov Fault on the Yenisei site. It was also confirmed that the zone of the dynamic impact of the major faults was less stable than the enclosing rock mass while the rock mass composed of dolerite dykes featured the highest stability.The authors appreciate participation of the researcher from the Nuclear Safety Institute, Russian Academy of Sciences, Candidate of Geological and Mineralogical Sciences O. A. Morozov in this study. This work was conducted in the framework of budgetary funding of the Geophysical Center of RAS, adopted by the Ministry of Science and Higher Education of the Russian Federation.The authors express their gratitude to the students of the NUST MISIS’ College of Mining A. V. Nikitenkova and A. K. Niyaz for the help in processing geological data.

Large scale Vertical SEALing (VSEAL) test: Impact of gas migration on bentonite based vertical seals

The Institute for Radiological protection and Nuclear Safety (IRSN) has launched the VSEAL project to investigate the impact of gas migration on the long-term performance of bentonite based vertical sealing systems. This project relies on in-situ experiments emplaced in IRSN´s Underground ResearchLaboratory (URL) in Tournemire (South France) and small-scale tests conducted in laboratory. The studied material is mixture of MX80 bentonite high density pellets and powder which are being evaluated as possible sealing materials in deep geological repositories. The test is still in progress, but the collected set of data provides already valuable information of the hydro mechanical behavior of Vertical sealing systems during hydration.

Combining short-range dispersion simulations with fine-scale meteorological ensembles: probabilistic indicators and evaluation during a 85Kr field campaign

Abstract. Numerical atmospheric dispersion models (ADMs) are used for predicting the health and environmental consequences of nuclear accidents in order to anticipate countermeasures necessary to protect the populations. However, these simulations suffer from significant uncertainties, arising in particular from input data: weather conditions and source term. Meteorological ensembles are already used operationally to characterize uncertainties in weather predictions. Combined with dispersion models, these ensembles produce different scenarios of radionuclide dispersion, called “members”, representative of the variety of possible forecasts. In this study, the fine-scale operational weather ensemble AROME-EPS (Applications of Research to Operations at Mesoscale-Ensemble Prediction System) from Météo-France is coupled with the Gaussian puff model pX developed by the IRSN (French Institute for Radiation Protection and Nuclear Safety). The source term data are provided at 10 min resolution by the Orano La Hague reprocessing plant (RP) that regularly discharges 85Kr during the spent nuclear fuel reprocessing process. In addition, a continuous measurement campaign of 85Kr air concentration was recently conducted by the Laboratory of Radioecology in Cherbourg (LRC) of the IRSN, within 20 km of the RP in the North-Cotentin peninsula, and is used for model evaluation. This paper presents a probabilistic approach to study the meteorological uncertainties in dispersion simulations at local and medium distances (2–20 km). First, the quality of AROME-EPS forecasts is confirmed by comparison with observations from both Météo-France and the IRSN. Then, the probabilistic performance of the atmospheric dispersion simulations was evaluated by comparison to the 85Kr measurements carried out during a period of 2 months, using two probabilistic scores: relative operating characteristic (ROC) curves and Peirce skill score (PSS). The sensitivity of dispersion results to the method used for the calculation of atmospheric stability and associated Gaussian dispersion standard deviations is also discussed. A desirable feature for a model used in emergency response is the ability to correctly predict exceedance of a given value (for instance, a dose guide level). When using an ensemble of simulations, the “decision threshold” is the number of members predicting an event above which this event should be considered probable. In the case of the 16-member dispersion ensemble used here, the optimal decision threshold was found to be 3 members, above which the ensemble better predicts the observed peaks than the deterministic simulation. These results highlight the added value of ensemble forecasts compared to a single deterministic one and their potential interest in the decision process during crisis situations.

Software Development for the Nuclear and Radiation Hazardous Objects Elimination Projects Cost Estimating Using Digital Modeling

This article discusses the problems of estimating the projects cost for decommissioning nuclear and radiation hazardous facilities based on BIM data. The software developed at Nuclear Safety Institute of the Russian Academy of Sciences for planning and analyzing decommissioning facilities at the pre-project stage processes is described. In particular, the software main functions are demonstrated: evaluation of the dismantling and decontamination works cost, forecasting the waste generated volume, technological processes planning and the safe waste management cost evaluation, results analysis taking into account the uncertainty of the initial data and sensitivity analysis. The scientific novelty consists in the development by a team of authors of software for financial and economic planning of decommissioning of nuclear and radiation hazardous facilities on the basis of digital information 3D models of objects being created. The need for such development is conditioned by the requirements for systematization and analysis of data on nuclear waste at the preparatory stage for the selection of effective technologies for dismantling and decontamination works and management of radioactive waste, as well as to improve the efficiency of individual decommissioning projects and the Federal Target Program "Providing Nuclear and Radiation Safety for 2016-2020 and for the period up to year 2035" generally.

Developing Approaches to Analysis of Melt Motion on the Surface of a Fuel Pin

Based on the analysis of experiments carried out on simulators of fuel elements with a cladding of low-melting metals, models of melting and movement of the melt over the surface of fuel elements are proposed. The presented work is a continuation of experimental and theoretical work carried out jointly at the Nuclear Safety Institute, Russian Academy of Sciences, and Kutateladze Institute of Thermophysics, Siberian Branch, Russian Academy of Sciences, to study the features of the movement of a melt over a cylindrical surface that simulates the surface of a fuel element. It analyzes the factors that determine the thermal destruction of fuel elements. One of the important conclusions drawn from the analysis of experimental data was the conclusion about the prevailing film laminar runoff in the heated part of the fuel element simulator. Using this assumption, an analytical model was built that allows predicting the mass of a fuel element carried out of its limits during melting. As a result of calculations using the proposed model, it was shown that, for an accident with the introduction of positive reactivity, when the power increase can reach ten ratings, a modification of the model is necessary to take into account the turbulent runoff regime. For this purpose, a simplified model of the flow of the melt was built. A comparison is made with known techniques and it is shown that the proposed approach makes it possible to describe the turbulent flow regime with satisfactory accuracy. The modified model made it possible to calculate the features of the movement of a stainless steel melt over the surface of a fuel element of a fast neutron (BN) reactor. Information was obtained on the thickness of the melt, its velocity and Reynolds number, depending on the conditions of drainage, and on the mass carried outside the core.

Concrete Modeling for Neutron Transport and Associated Sensitivity Studies Performed at the AMANDE-MIRCOM Facility

Concrete plays a major role in nuclear facilities as protection against radiation. However, its chemical composition, which is fundamental, is often unknown. Several concrete samples, extracted from the AMANDE-MIRCOM Institute for Radiological Protection and Nuclear Safety (IRSN) facility, were analyzed. Various simulations were performed in order to evaluate the neutron fluence behind a 40-cm-thick concrete wall. These simulations were compared to experimental measurements performed with a Bonner sphere spectrometer and a neutron survey meter. No set of parameters tested was able to produce a simulation accurately matching all the experimental results, but sensitivity studies on several parameters highlight that the three most sensitive parameters are the hydrogen content, the density, and the concrete inhomogeneity. To improve the agreement between the simulations and the measurements, the concrete inhomogeneity modeling should be studied further. Nevertheless, using concrete compositions that are close to reality, especially for hydrogen content, is crucial to correctly simulate neutron transport.

Радиационный риск и вероятностный анализ безопасности третьего уровня объектов использования атомной энергии

The paper considers the current state in the field of level 3 probabilistic safety assessment (PSA Level 3) of nuclear facilities. The issues on rationing individual and social risks in various indus-tries and practice of legislative control in some countries are examined. Some attention is paid to the identification of acceptable risk level and features of radiation risk perception. The paper ad-dresses possibilities of use of PSA Level 3 methodology for radiation risk assessment. We pro-pose a practical approach to quantify specific features that may affect individual, collective and social radiation-associated cancer risks at the initial stage of severe radiation accidents at nuclear power plants. PSA Level 3 capacities are demonstrated on the example of individual and social radiation risks assessment at the initial stage of a severe accident at a model nuclear power plant. When performing numerical simulations of the consequences of the accident of such type with the program developed at the Nuclear Safety Institute of the Russian Academy of Sciences, the nuclide composition of the emission is similar to one of the emissions during the Fukushima accident, and the distribution of population density and weather conditions correspond to real values in Fukushima Prefecture. We evaluated the effectiveness of various options for protective countermeasures based on the results of risk assessments. We have shown that the PSA Level 3 methodology allows us to obtain information about the risks associated with accidents at nuclear facilities in the form of standard risk-informative safety management methods. This will strength-en scientific and technical support for the processes of interaction with experts, decision-makers, the media and the public in making a wide range of decisions in the event of a threat, occurrence and elimination of the consequences of an emergency with a radiation factor.

Transient heat exchanges under fast Reactivity-Initiated Accident

Heat exchanges during fast transient are very complex phenomena. Many studies have been led in this field, trying to quantify the heat exchanges between a heating wall, under fast increasing power, and a coolant. This kind of situation is encountered for instance during RIA (Reactivity-Initiated Accidents) in nuclear reactors, whose characteristics times of wall heat flux excursion can be as low as orders of 1 ms. The CABRI reactor, an experimental pulse reactor funded by the french Institute for Radiological protection and Nuclear Safety (IRSN) and operated by CEA at the Cadarache nuclear center (France), was build in order to study and thus to better understand RIA effects on nuclear fuels. Heat exchanges characterization is definitely one of major points to tackle in the understanding and modeling of such transients, involving single and two phase flows.This paper broaches the question of single phase heat exchanges coefficients during fast transient power excursions in the CABRI reactor.The single phase pure convection is the overlap of two main mechanisms: the advection (wall axial supply in upstream cold water) and the turbulent mixing inside the boundary layer. This latter phenomenon is due to vortices in boundary layers inducing radial mixing through the velocity boundary layer. In most cases involving turbulent flows, the turbulent mixing phenomenon is preponderant. To these mechanisms is added a pure conduction heat exchange mechanism through the thermal boundary layer. Convection and conduction overlap defining a conducto-convective heat transfer coefficient, improperly only called “convection” coefficient. This paper suggests a way to model this coefficient during a CABRI-RIA transient by a non-linear superposition of transient pure conduction and convection mechanisms. Two distinct transient phases are considered; the first one presenting a wall heat flux of exponential shape followed by the second one during which the wall heat flux remains constant. The potency of this analytical model, which is implementable into computational system tools, is demonstrated.

Multiphysics CATHARE2 modeling and experimental validation methodology against CABRI transients

CABRI is an experimental pulse reactor, funded by the french Institute for Radiological protection and Nuclear Safety (IRSN) and operated by CEA at the Cadarache nuclear center, France. Its aim is to study and thus to better understand RIA (Reactivity-Initiated Accident) effects on nuclear fuels.The restart of the CABRI reactor in 2015 offers the opportunity to validate tools involving multiphysic calculation schemes on reactivity insertions (RI) transients at a system scale. Physics of the in-pile CABRI tests is complex and implies various physical fields (solid and fluid thermics, thermal-hydraulics, mechanics and neutronics) and their coupled effects. The reactivity insertion in CABRI is mastered by the depressurization of a neutron absorber (3He), contained into transient rods. The increase of the neutron flux inside the core induces the neutronic power increase inside the core and a multiphysic effect, called “TOP” effect, inside the in-core transient rods that causes an acceleration of the gas depressurization.The challenge is thus to manage the simulation of these RIA complex transients catching the governing multiphysical phenomena to be identified. In order to achieve this objective, a model for the CABRI transients and a scientific calculation tool have been developed and should be validated. This paper proposes a modeling of governing multiphysic phenomena of RIA transients in CABRI reactor based on a Quantified Phenomena Identification and Ranking Table (QPIRT). These models are introduced in the CATHARE2 tool in a dedicated version named PALANTIR and Best-Estimate simulation results are compared with experimental data obtained on CABRI commission tests: core power and 3He pressure evolution.

Potential of Surface-to-Tunnel Seismic Tomography to Detect Vertical Faults: Application to the Tournemire Underground Research Laboratory, France

The Tournemire Underground Research Laboratory (Aveyron, France), developed by the French Institute for Radiological Protection and Nuclear Safety, is composed of a tunnel excavated in an argillaceous geological layer belonging to a limestone–argillite–limestone sub-horizontal sedimentary sequence. Sub-vertical fault zones are intercepted by drifts and boreholes excavated from the tunnel in the argillaceous layer at a depth of about 250 m. These fault zones are characterized by velocities lower than those of the embedding rocks. In this study, we assess the capacity of first arrival traveltime tomography to detect and characterize such fault zones in a surface-to-depth transmission configuration with a limited angle aperture. The source at the surface is a sledgehammer and turns out to be very repetitive. The source trigger is transmitted to the acquisition bay via a VHF transmitter. Receivers are located at depth in underground works. First arrivals are clear and traveltime picking is straightforward. Before inverting real data, we conduct synthetic tests with the real acquisition geometry to assess limits and validity of first arrival traveltime tomography to detect a lower velocity fault zone using different initial configurations and inversion constraints during the inversion. We show that introducing a priori information on velocities in the shallow subsurface, limestones and argillaceous layers and constraints on maximum velocities greatly helps improving results and subsequent interpretation. Velocities obtained from the real data display a main lower velocity zone that extends from the galleries in the argillaceous layer right up to the surface in limestones. The location and extension of this lower velocity zone are consistent with fault and fracture zones mapped from geological observations.

Berkut Code Validation on Post-Reactor Studies of Irradiated Bn-600 Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel

An advanced version of the BERKUT-U/V2.1 code intended for modeling the behavior of fuel rods with oxide or nitride fuel in standard and emergency operating regimes of fast reactors with liquid-metal coolant is being developed at the Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS), as part of the project Next-Generation Codes of Project Proryv (Breakthrough). The present work reports the results of the validation of an advanced version of the code based on post-reactor studies of BN-600 fuel rods (KETVS-1, -6) and BREST-OD-300 fuel rods (ETVS-5) with mixed uranium-plutonium nitride fuel, irradiated in the BN-600 reactor.

SciFi detector and associated method for real-time determination of profile and output factor for small fields in stereotactic radiotherapy.

PurposeFor determining small‐field profile and output factor during stereotactic radiotherapy quality assurance (QA) procedures, we propose a novel system based on the scintillating fiber (SciFi) detector with output image acquisition and processing to allow real‐time monitoring of profile and output factor.Materials and methodsThe employed detector is a SciFi detector made of tissue‐equivalent scintillating plastic fibers arranged in 6‐layer fiber ribbons with a fiber pitch of 275 μm in each layer. The scintillating signal at the detector output is acquired by a sCMOS (scientific complementary metal–oxide–semiconductor) camera and represents the projected field profile along the fibers axis. An iterative reconstruction method of the field from its projected profile based on a priori knowledge of some features of the radiation field defined by the stereotactic cones is suggested. The detector with implemented data processing has been tested in clinical conditions, for determining beam profiles and output factors, using cone collimators of different sizes from 4 to 15 mm diameter. The detector under test was placed at 1.4 cm depth and 98.6 cm source to surface distance (SSD) in a water‐equivalent phantom and irradiated by a 6 MV photon beam.ResultsThe reconstructed field profiles obtained from the detector are coherent with data from EBT3 radiochromic films, with differences within ±0.32 mm for both the FWHM and the penumbra region. For real‐time determination of the field output factor, the measured data are also in good agreement with data independently determined by the French Institute for Radiological Protection and Nuclear Safety (IRSN) based on radiochromic films and thermoluminescent 1 × 1 mm2 micro‐cubes dosimeters (TLD). The differences are within ±1.6% for all the tested cone sizes.ConclusionsWe propose and have tested a SciFi plastic scintillating detector with an optimized signal processing method to characterize small fields defined by cone collimators. It allows the determination of key field parameters such as full width at half maximum (FWHM) and field output factors. The results are consistent with those independently measured using TLD and radiochromic films. As the SciFi detector does not require a correction factor, it is in line with the International Atomic Energy Agency (IAEA) and the American Association of Physicists in Medicine (AAPM) TRS‐483 recommendations, and can be suitable for online QA of small radiation fields used in photon beam radiotherapy, and is compatible with MRI‐LINAC.

Best Estimate Plus Uncertainty (BEPU): Why It Is Still Not Widely Used

As a technical support organization for French Public Authorities, the Institute for Radioprotection and Nuclear Safety (IRSN) shall participate in policy making on nuclear safety and radiation protection. In the framework of its current activity, the IRSN performs analysis and comparison of different safety assessment approaches, including Best Estimate Plus Uncertainty (BEPU), among others. Although BEPU is not yet recognized worldwide, the authors consider it a promising way to stimulate the involvement of a new generation of numerical tools and high-fidelity experiment data for nuclear safety assessment and development of nuclear technology. This paper discusses enabling factors and constraints for BEPU from technical and global points of view, relying on scientific background as well as current practice with respect to its potential role in regulation and, more generally, in a decision-making process.

Nuclear data and applications at the nuclear safety and radioprotection institute: Analysis, evaluation and application

The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is the French public service expert in nuclear and radiation risks with activities that cover all the related scientific and technical issues. IRSN provides expertise in support of radiation protection and nuclear safety risks. A great deal of research work carried out at IRSN concerns the improvement of computational tools and data used on a daily basis by nuclear safety assessment experts. Particularly in this paper the approach used to improve data quality and its uncertainties is presented. Computation tools and nuclear data evaluation methodology are discussed.

Verification of the EUCLID/V2 Code Based on Experiments Involving Destruction of a Liquid Metal Cooled Reactor’s Core Components

The article presents the results obtained from verification of the EUCLID/V2 coupled code developed at the Nuclear Safety Institute of the Russian Academy of Sciences, which is intended for analysis of accident conditions in liquid metal cooled fast reactors. The additional capabilities available in the code in comparison with its first version include, in particular, analysis of individual equipment components in the 3D approximation, consideration of the transport of fission products and corrosion products in the coolant and in the nuclear power plant buildings, and also analysis of severe accidents in a fast reactor. The article presents the code verification results and assessment of its applicability for analysis of accidents involving destruction of fuel pins and the reactor core. The verification was carried out against the data obtained at experimental facilities and from analytic tests. Information about the key experiments used to validate the code is briefly outlined. In particular, data of experiments carried at the Oak Ridge, Argonne, and Sandia National Laboratory, the United States; at the National Nuclear Center of the Kazakhstan Republic; and on the test bench at the Nizhny Novgorod State Technical University (NSTU) in Russia are used. Modules of the coupled code EUCLID/V2 integrated code verification matrix are given. The errors of calculating the parameters most important for analysis of an accident’s consequences evaluated using the EUCLID/V2 code are proven with due regard to the computation and experimental results. The ranges of parameters within which the code has been verified are determined. The uncertainty and sensitivity of computation results are also analyzed based on the results from simulating certain experiments. Factors having the main influence on the computation results are determined. It is shown that the computation results are consistent with the experimental results subject to the input data uncertainty.