Computer Code System Research Articles

Development and Assessment of U.S. Nuclear Regulatory Commission Thermal-Hydraulic System Computer Codes

A review is provided of the reasons why the US Nuclear Regulatory Commission needs thermal-hydraulic system computer codes, the assumptions and approximations contained within these codes, and the reasons why test data are required to assess the accuracy of the codes. Specific examples of codes and test programs are given. The use of computer codes assessed against data from scaled test facilities to predict the full-scale plant response is discussed. A method to help focus resources and the need for quantifying code uncertainties are discussed. This paper concentrates on the loss-of-coolant accident (LOCA) because most of the analytical and experimental research has been concentrated in LOCAs.

Computer Code System DSOCEAN for Assessing the Collective Dose of Japanese due to Radionuclides Released to the Ocean from a Reprocessing Plant

A computer code system DSOCEAN has been developed for assessing the collective dose of Japanese due to radionuclides released to the ocean from a spent nuclear fuel reprocessing plant. This computer code system uses a box model which represents the transfer of radionuclides between boxes of seawater into which the ocean around Japan is divided. The code system consists of a series of three interlinked main computer codes, which estimates the exchange rates of radionuclides between the boxes, the radionuclide concentrations in each box, and the collective dose from various exposure pathways, respectively.By using DSOCEAN, two calculations were carried out to estimate the collective dose from a liquid effluent. One is associated with a routine release of radionuclides from a hypothetical reprocessing plant. The other is an application of the code system to disposal of liquid radioactive waste to the surface water of the ocean. The calculated results identified important radionuclides and exposure pathways.

Preliminary Accident Analysis to Support a Passive Depressurization System Design

The new generation of evolutionary nuclear power plants, e.g., the Westinghouse AP600 and the General Electric simplified boiling water reactor, relies on a full reactor coolant system (RCS) depressurization to allow gravity injection from an in-containment tank and thereby assure long-term core cooling. Studies performed to support the licensing process and design of both evolutionary and innovative reactors have shown that cold water injection may, under particular plant conditions, induce a large plant depressurization.Preliminary studies have been performed to support the design of a passive injection and depressurization system (PIDS) based on the idea of depressurizing the RCS by mixing cold water with the RCS hot water and inducing steam condensation in the primary system. The analyses, performed with the RELAP5/MOD3 computer code, show the response of a typical midsize pressurized water reactor plant [two loops, 600 MW(electric)] equipped with the PIDS.Different RCS injection locations including pressurizer, vessel upper head, and hot leg, and actuation at different residual reactor coolant masses have been investigated. Several factors, including RCS mixing, RCS residual mass at PIDS actuation, PIDS injection flow rate, and steam generator conditions, have been shown to affect the plant depressurization.The PIDS performance has also been verified against the following reference severe accident scenarios: (a) complete station blackout event, and (b) a small-break loss-of-coolant accident and concomitant station blackout event.Preliminary experimental activities to support the PIDS concept have already been performed. Additional experimental activities, including integrated system tests, have been planned to support system development and computer code validation.

Computer Code System DSOCEAN for Assessing the Collective Dose of Japanese due to Radionuclides Released to the Ocean from a Reprocessing Plant.

A computer code system DSOCEAN has been developed for assessing the collective dose of Japanese due to radionuclides released to the ocean from a spent nuclear fuel reprocessing plant. This computer code system uses a box model which represents the transfer of radionuclides between boxes of seawater into which the ocean around Japan is divided. The code system consists of a series of three interlinked main computer codes, which estimates the exchange rates of radionuclides between the boxes, the radionuclide concentrations in each box, and the collective dose from various exposure pathways, respectively.By using DSOCEAN, two calculations were carried out to estimate the collective dose from a liquid effluent. One is associated with a routine release of radionuclides from a hypothetical reprocessing plant. The other is an application of the code system to disposal of liquid radioactive waste to the surface water of the ocean. The calculated results identified important radionuclides and exposure pathways.

96/00314 Study of the structure and components of tar from Shenfu coal

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.

Development of a nuclear-thermal coupled calculation code system

Development of a nuclear-thermal coupled calculation code system

Development of a nuclear-thermal coupled calculation code system

To obtain detailed time-dependent temperature distributions in fusion reactor nuclear components, a nuclear-thermal coupled calculation code system is developed. The code system couples the two-dimensional radiation transport code dot -3.5 and a finite element non-linear temperature analysis code adinat . The neutron flux and gamma-ray flux distributions in a fusion reactor model are first calculated with the dot -3.5 code, and these are then converted to time-dependent nuclear heating rate distributions by the apple -3 code. The nuclear heating rate distribution is then converted to a data format compatible with the input for the adinat code, to calculate the temperature distribution. The coupled calculation code system, i.e. dot-apple-adinat , has been tested for the calculation of the time-dependent temperature distributions in graphite, molybdenum and Li 2 CO 3 probes for nuclear heating measurements in the fusion neutronics course at the Japan Atomic Energy Research Institute.

Transmutation of the actinide neptunium-237 with a hybrid reactor

Calculations and analyses on the transmutation of long-lived actinide 237Np radioactive waste with a fusion-fission hybrid reactor are reported. A new waste management concept of burning HLW (high-level waste), utilizing the released energy and converting 237Np into 239Pu fissile fuel through transmutation has been adopted. Time-dependent depletion calculations were performed using a modified version of the BISON 1.5 one-dimensional neutron transport and burn-up calculation code system. The transmutation rate of 237Np and its relationship to neutron wall loading and Pu and Np concentrations in the transmutation zone were explored; relevant results are given.

Methodology of Safety Assessment and Sensitivity Analysis for Geologic Disposal of High-Level Radioactive Waste

A deterministic safety assessment methodology has been developed to evaluate long-tern1 radiological consequences associated with geologic disposal of high-level radioactive waste, and to demonstrate a generic feasibility of geologic disposal. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. A computer code system GSRW thus developed is based on a non site-specific model, and consists of a set of sub-modules for calculating the release of radionuclides from engineered barriers, the transport of radionuclides in and through the geo-sphere, the behavior of radionuclides in the biosphere, and radiation exposures of the public. In order to identify the important parameters of the assessment models, an automated procedure for sensitivity analysis based on the Differential Algebra method has been developed to apply to the GSRW.

Methodology of Safety Assessment and Sensitivity Analysis for Geologic Disposal of High-Level Radioactive Waste.

A deterministic safety assessment methodology has been developed to evaluate long-term radiological consequences associated with geologic disposal of high-level radioactive waste, and to demonstrate a generic feasibility of geologic disposal. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. A computer code system GSRW thus developed is based on a non site-specific model, and consists of a set of sub-modules for calculating the release of radionuclides from engineered barriers, the transport of radionuclides in and through the geo- sphere, the behavior of radionuclides in the biosphere, and radiation exposures of the public. In order to identify the important parameters of the assessment models, an automated procedure for sensitivity analysis based on the Differential Algebra method has been developed to apply to the GSRW.

Evaluation of core thermal and hydraulic characteristics of HTTR

Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88% and 1492°C, respectively.

AUTOLOAD, An Automatic Optimal Pressurized Water Reactor Reload Design System with an Expert Module

An automatic optimal pressurized water reactor (PWR) reload design expert system AUTOLOAD has been developed. It employs two important new techniques. The first is a new loading priority scheme that defines the optimal placement of the fuel in the core that has the maximum end-of-cycle state k[sub eff]. The second is a new power-shape-driven progressive iteration method for automatically determining the burnable poison (BP) loading in the fresh fuel assemblies. The Haling power distribution is used in converting the theoretically optimal solution into the practical design, which meets the design constraints for the given fuel assemblies. AUTOLOAD is a combination of C and FORTRAN languages. It requires only the required cycle length, the maximum peak normalized power, the BP type, the number of fresh fuel assemblies, the assembly burnup, and BP histories of the available fuel assemblies as its input. Knowledge-based modules have been built into the expert system computer code to perform all of the tasks involved in reloading a PWR. AUTOLOAD takes only [approximately] 30 CPU min on an IBM 3090 600s mainframe to accomplish a practical reload design. A maximum of 12.5% fresh fuel enrichment saving is observed compared with the core used by the utility.

Development of Worldwide Version of System for Prediction of Environmental Emergency Dose Information: WSPEEDI, (III)

The numerical models of WSPEEDI (Worldwide version of System for Prediction of Environmental Emergency Dose Information) were revised. The new models are formulated with a terrain following vertical coordinates with the compressibility of the atmosphere being considered. The models have been integrated into a computer code system together with system control software, worldwide geographic database, meteorological data pre-processor and graphics. The performance of the code system has been evaluated against the Chernobyl case with reliable source terms, well established meteorological input and a comprehensive monitoring database, which were provided in ATMES (Atmospheric Transport Model Evaluation Study) coordinated by IAEA, WMO and CEC.

Criticality Data and Validation Studies of Mixed-Oxide Fuel Pin Arrays in Pu+U+Gd Nitrate

Critical experiments were performed at the Pacific Northwest Laboratory's Critical Mass Laboratory in 1987 and 1988 with a heterogeneous array of mixed-oxide (MOX) fuel pins immersed in mixed plutonium-uranium nitrate solutions. The 996 fuel pins, on a 1.40-cm-square pitch, were configured in a cylindrical array. The solution heavy metal concentrations ranged from 4 to 468 g/[ell] and had a Pu/Pu + U ratio of 0.2. Critical experiments were also performed with gadolinium added to the fissile solution. These experiments were designed to simulate conditions in a MOX fuel dissolver, where fuel lumps are moderated by aqueous solutions containing fissile nuclides, with and without a soluble neutron poison. For the experimental conditions examined, it was determined that the critical size of the system increased as the heavy metal concentration increased. The criticality data were used to validate two versions of the scale computer code system and the 27-energy-group cross-section library, derived from the Evaluated Nuclear Data File B Version IV. The calculations results indicate that SCALE-2 has some difficulty in modeling these systems. Modifications in SCALE-4 have led to more accurate K[sub eff] results.

Development of Worldwide Version of System for Prediction of Environmental Emergency Dose Information: WSPEEDI, (III). Revised Numerical Models, Integrated Software Environment and Verification.

The numerical models of WSPEEDI (Worldwide version of System for Prediction of Environmental Emergency Dose Information) were revised. The new models are formulated with a terrain following vertical coordinates with the compressibility of the atmosphere being considered. The models have been integrated into a computer code system together with system control software, worldwide geographic database, meteorological data pre-processor and graphics. The performance of the code system has been evaluated against the Chernobyl case with reliable source terms, well established meteorological input and a comprehensive monitoring database, which were provided in ATMES (Atmospheric Transport Model Evaluation Study) coordinated by IAEA, WMO and CEC.

Two-phase flow instrumentation

A careful measurement of the relevant two-phase flow parameters is the basis for the understanding of many thermohydraulic processes. Reliable two-phase instrumentation is therefore essential for the connection between analysis and experiment especially in the nuclear safety research where accident scenarios have to be simulated in experimental facilities and predicted by complex computer code systems. Ambitious development programs have been carried out in many institutions and countries to improve two-phase instrumentation. Advantages as well as limitations of some of these systems are discussed in this paper. In the last 10 to 15 years substantial progress has been made, although there are still goals for further developments. And it is still a fact that in many cases measured data taken from large experimental facilities cannot be compared directly to the parameters calculated by the codes. Careful comparison and interpretation of both calculated and measured results by experienced researchers will continue to be the key for thermohydraulic understanding of complex two-phase phenomena in the future.

MAGS: A computer code system to analyze the 3D quench propagation in forced flow cooled superconducting magnet systems

For the detailed analysis of the propagation of normal conducting zones in forced flow cooled superconducting coils the code system MAGS has been developed at Kernforschungszentrum Karlsruhe. The code system is modular, i.e., depending on the desired degree of details of the results the user can choose between different modules and options to adjust the analysis effort to the requirements. The paper gives a rough description of the modules. An example for application is the analysis of a quench in an ITER coil using a detailed and a coarse model. It turned out that the coarse model gives results with an error of 50% in peak temperature compared to the detailed results. Keeping in mind that the computing effort is reduced with the coarse model by two orders of magnitude the coarse model is an interesting option of MAGS.

A new algorithm for EGS4 low-energy electron transport to account for the change in discrete interaction cross-section with energy

This paper presents a new algorithm for the EGS4 Monte Carlo computer code system to account for the change in electron discrete interaction cross-section as a function of energy for low-energy electron transport. The current EGS4 algorithm uses a crude method to deal with this change which is only valid for high-energy cutoffs (AE > 560 keV) and therefore results in a systematic error in low-energy electron transport. In the new algorithm a variable electron mean free path correction factor is used to account for the effect of the overestimated electron mean free path at the beginning of an electron path. The electron path length is then more accurately sampled from the averaged electron discrete interaction cross-section along the path using a linear variation model. Calculations of the absorbed dose and particle fluence in different geometries have been carried out. The results show that the number of discrete interactions for low-energy electron transport (AE = 512 keV) increases by more than 20% with the new algorithm compared with the current EGS4 algorithm. The depth-dose calculations show a 1–2% difference around the depth of maximum dose for electron energies of 10 keV to 1 MeV. The difference in the electron fluence obtained with the new algorithm and that with the current algorithm is about 4% at energies of 100–300 keV. It is recommended that the new algorithm should be implemented in the EGS4 system for the calculations of the absolute dose or fluence with electron energies below 1 MeV and energy cutoff AE smaller than 531 keV.

ESTIMATING THE COSTS FOR JAPAN’S JPDR PROJECT

There has been an international flavor of cooperation in the commercial development of nuclear power. This cooperation is most strongly established with the development of strategies for decommissioning. Many countries are beginning this phase at about the same time; Japan, Canada, Germany, and the United States all are involved in dismantlement projects. This chapter, along with the next four chapters, addresses various aspects of the methodological approaches being developed to estimate decommissioning costs. Each country has adopted a different strategy for reactor decommissioning, taking its political and technical situations into consideration. Japan also has its own strategy of reactor decommissioning, reflecting the geographic and economic positions peculiar to Japan. The Japan Power Demonstration Reactor (JPDR) decommissioning program is in the process of establishing a decommissioning data base and a cost estimation methodology, as well as developing new technology for reactor decommissioning. Various information about the JPDR dismantling has been accumulated in the decommissioning data base, which will be used for: (1) planning future decommissioning of commercial nuclear power reactors; (2) verifying the developed code system for management of reactor decommissioning; and (3) managing the ongoing JPDR dismantling. The computer code system developed in this program is expected to contribute to studying cost estimation and the optimization of decommissioning plans for commercial nuclear power reactors.

Theoretical Background of the Dosimetry System 86 (DS86)

DS86 is a calculation code system for the estimation of the radiation dose of atomic bomb survivors at Hiroshima and Nagasaki. The DS86 calculation process is as follows: 1. Compute free-field radiation environment: radiation transport calculation can be done for the free air field at Hiroshima and Nagasaki. 2. Compute shielded radiation environment: depending on each shielding situation, the shielding can be calculated by using either the Nine-Shielding Parameter or the Globe Shielding Data. 3. Compute radiation environment at the organ site: organ kerma and fluence spectra can be calculated using a transport calculation code depending on each survivor sex, age, orientation and posture.