External radiation of a container used for dry storage of spent VVER-1000 nuclear fuel from the Zaporozhie nuclear power plant

The program MCNP (Monte Carlo Modeling of radiation transfer) is used to calculate the characteristics of external neutron and γ radiation from a ventilated dry-storage container for spent nuclear fuel. Data are obtained on the spatial, energy, and angular distribution of the neutron and γ-ray flux outside the container and the dependence of the dose rate on the storage time of the spent fuel is determined. It is shown that γ-rays make the main contribution to the dose rate on the side surface of the container and neutrons do on the cover. The computed dose rate is 1.4 times higher than the measured value on an individual loaded container at the Zaporozhie nuclear power plant.     

Criticality safety of the ET-RR-1 new spent fuel storage pool

A new ET-RR-1 spent fuel storage pool is now under construction on the reactor site at Inshass. In addition, the pool is designed to accommodate spent fuel of MTR type as well. Criticality safety of this pool for the different fuel types has been evaluated as a function of U²³⁵ loading. The effect of fuel element separation (rows and columns) on the eigenvalue has been studied. As a conservative assumption, the pool is assumed to be filled with fresh fuel. The eigenvalue considering a realistic degree of fuel burn-up was determined in order to determine the safety margin. The calculations have been carried out using the code packages of the National Center for Nuclear Safety and Radiation Control.     

Heat- and neutron-physical characteristics of power-generating units of nuclear power plants with VVER-1000 during operation above nominal power

The margin to critical level of heat exchange of thermal power and coolant temperature at the core exit as well as the nonuniformities of the energy release under the operating conditions of a nuclear power plant with VVER-1000 above nominal power is analyzed for the No. 2 unit of the Balakovo nuclear power plant. It is confirmed that the safety criteria comply with OPB-88/97. The content of each of the three main stages of operation is presented. Conclusions are drawn and recommendations are made concerning updating the technical design of the reactor facility in order to increase the power level.     

Thermal analysis of the heat exchanger for a spent nuclear fuel storage pool

A numerical solution is provided to predict the transient temperature distribution of both fluids in the U-tube heat exchanger of a spent nuclear fuel storage pool. A finite element method, with the Galerkin approach, is used to solve the set of five partial differential equations of energy conservations, with arbitrary inlet and boundary conditions. The results are obtained with very low computation time, through a computer program on a CDC 730, which can be easily linked to other thermal hydraulic codes for the storage pool.To show the capabilities of the program, some results are presented, concerning step response and other transient operations of the exchanger.The validation of the method has been performed comparing the numerical results with the exact steady state analytical solutions available in literature; the agreement is very satisfactory.     

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.     

Computational and experimental validation of protective features of transport packages for long-term storage of spent nuclear fuel from nuclear power plants

Abstract

The purpose of this paper is to describe the work of our Institute (RFNC-VNIITF) on the development of a unique long transport container using new manufacturing technologies. The container is designed for transport and long-term storage of spent fuel elements more than 10 m long. The disposal of such elements usually implies their cutting. This requires considerable expense in building special cutting plants. Another way is to use long protective shells to contain and store the entire fuel cell. Development of such containers has been made possible by the use of rolled-strip technology ensuring maximum strength of containers. The paper contains certain results of calculations and experimental research justifying and confirming the protective properties of the long package. To confirm reliability with minimum material consumption an available package model, manufactured by new technology, is used. This package was enhanced in the flange area to simulate strictly a flange joint of a full-scale transport package. Calculations and experimental work were performed according to IAEA regulations. During investigations the full-scale cask was considered as both a model and a prototype as it simulated completely the bottom and the flange joint 'lid-cask body' (i.e. being a model) while being shorter and thinner than the cylindrical part of the body (i.e. being a prototype). Here it is called simply a prototype. The complete simulation was not set as an objective. At the same time the cask body loading for tests simulating an air crash was done with regard to the full-scale design. The paper contains results obtained analytically on the basis of experimental data to justify the properties of a full-scale package without experimentally testing it: therefore only some photographs and experimental results for the prototype are presented here. The title of the paper indicates that the analytical study was conducted with use of numerical constants obtained experimentally. It is impossible to present all the information, including experimental research, here because of the large scope of the work performed and the limited space available.     

Nuclear criticality safety considerations for CASTOR casks with spent fuel of the Rossendorf nuclear devices

The spent fuel of the shut down Rossendorf nuclear devices is to be loaded into storage and transport casks of the type CASTOR-MTR-2. According to the variety of different nuclear devices at the Rossendorf site, the Rossendorf fuel is characterized by a great variety with regard to geometry, material, enrichment, and burn-up. According to the special loading conception, the fuel is embedded in aluminium bodies that the fill the CASTOR. The void fraction within the CASTOR is very small resulting in a small water fraction if water flooding is assumed. The criticality safety is proved by MCNP and OMEGA calculations. These are independent codes that use a completely different data base. The results of both codes agree very well demonstrating the reliability of the calculations. Apart from the proof of criticality safety, some interesting features were found mainly as a result of the very small water fraction.     

Heat-treatment and storage technology for spent ion-exchange resins for new-generation nuclear power plants

A technology for thermovacuum drying of spent radioactively contaminated ion-exchange resins was developed. It was shown experimentally that thermovacuum drying yields a product suitable for long-term storage and subsequent disposal. Recommendations were made for using NZK and KMZ concrete and metal casks for holding dry, spent, low- and medium-activity ion-exchange resins without additional solidification. Because the volume of the product obtained is greatly reduced, thermovacuum drying will decrease by at least 15-fold (from 100 to 7) the number of casks required for long-term storage and subsequent disposal of spent ion-exchange resins produced in one year of operation of a single VVER-1200 power-generating unit.     

Validation of nuclear and radiation safety of a container for spent amb reactor fuel assemblies at the Beloyarskaya nuclear power plant

An approach is proposed for validating the nuclear and radiation safety of a container for spent fuel assemblies from AMB-100 and-200 reactors at the Beloyarskaya nuclear power plant. To validate the radiation safety, the characteristics of fuel assemblies and their classification according to the average fuel burnup in the casing, and the intensities of n and γ radiation in the casing are analyzed. Nuclear safety is validated on the basis of the concept of a “model” casing. This model makes it possible to obtain an upper estimate of the effective coefficient of neutron multiplication for all real casings with fuel assemblies. Calculations are used to determine the minimum necessary thickness of the vessel, bottom, and cover for 17-and 35-place casings. It is shown that no special neutron protection is needed. The container design to be developed meets the IAEA and OPBZ-83 safety standards. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 423–428, June, 2006.     

Investigation of xenon transient processes in VVER-1000 at the Tianwan nuclear power plant (China)

The integral and spatial xenon transient processes in the No. 1 unit of the Tianwan nuclear power plant (China) have been studied experimentally. A measurement method which is unconventional for VVER-1000 was tested in the investigations of the integral processes: the course of the xenon process was recorded according to the variation of the critical concentration of boric acid in the reactor at the same time as the concentration was calculated in real-time. The spatial transient processes were studied for the diametric and axial free xenon oscillations of the energy release in the core. It was confirmed experimentally that axial deformations of the energy release affect the power of the reactor as well as the associated operational particularities of the automatic power regulator. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 183–190, October, 2008.     

Choice of organic diluents for the extractive regeneration of the spent fuel of nuclear power plants

Conclusions The length of the n-alkane chain which provides a basis for the diluent can lie within the interval C₁₁-C₁₅. Within this interval, the ratio of the separate n-alkanes in the diluent can be regulated. It is better, however, to use the hydrocarbons C₁₁-C₁₅; their rather high flash point gives them a slight edge over the higher members of the group in regard to their compatibility with the extracting solvates of the actinide nitrates and also with regard to their hydrodynamic characteristics. The content of aliphatic acids and alcohols 0.01 M; of unsaturated compounds 0.005 M; of aromatic hydrocarbons 1 vol.%. These requirements may change in the future as more investigations are made and a deeper study is made of the factors which affect the behavior of the diluent in the extractive cycle.Translated from Atomnaya Énergiya, Vol. 47, No. 2, pp. 75–79, August, 1979.     

Pulsed pneumatic transport of solid wastes from reprocessing of spent nuclear power plant fuel to storage

The results of experimental investigations of the effect of a gel-like residue on the transport of pieces of structural materials of fuel assemblies from nuclear power plants by a pulsed pneumatic transport system to storage are presented. The data obtained show that even increasing the viscosity of the wetting medium substantially (by a factor of 100) has little effect on the transport regime and technology. 3 figures, 2 tables, 4 references. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 48–51, January, 2000.     

Hydrogen production, storage, and use at nuclear power plants

The present status of the problems of safe storage and use of hydrogen in the world hydrogen-energy sector are analyzed. Specific examples of foreign and domestic designs of atomic-commercial complexes based on operating nuclear power plants, viewed as hydrogen producers and users, are presented. A method of producing hydrogen accumulators with cartridges which contain microcapsules or capillaries, made of high-strength materials and filled with hydrogen under high presure (above 100 MPa), is proposed as a promising direction for solving storage and use problems. The mechanisms of introducing/extracting hydrogen into/from microelements in the space of the accumulators up to a working pressure of 0.2–1 MPa are based on diffusion and active thermomechanical principles. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 420–426, December, 2006.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

Interim storage of power reactor spent nuclear fuel (SNF) and its potential application to SNF separations and closed fuel cycle

Interim, centralized, engineered (dry cask) storage facilities for USA light water power reactor spent nuclear fuel (SNF) should be implemented to complement and to offer much needed flexibility while the Nuclear Regulatory Commission is funded to complete its evaluation of the Yucca Mountain License and to subject it to public hearings. The interim sites should use the credo reproduced in Table 1 [Bunn, M., 2001. Interim Storage of Spent Nuclear Fuel. Harvard University and University of Tokyo] and involve both the industry and government. The sites will help settle the 50 pending lawsuits against the government and the $11 billion of potential additional liabilities for SNF delay damages if Yucca Mountain does not being operation in 2020 [DOE, 2008a. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Stations (December)].Under the developing consensus to proceed with closed fuel cycles, it will be necessary to develop SNF separation facilities with stringent requirements upon separation processes and upon generation of only highly resistant waste forms. The location of such facilities at the interim storage sites would offer great benefits to those sites and assure their long term viability by returning them to their original status.The switch from once-through to closed fuel cycle will require extensive time and development work as illustrated in “The Path to Sustainable Nuclear Energy” [DOE, 2005. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles. DOE (September)]. A carefully crafted long term program, funded for at least 5 years, managed by a strong joint government–industry team, and subjected to regular independent reviews should be considered to assure the program stability and success.The new uncertainty about Yucca Mountain role raises two key issues: (a) what to do with the weapons and other high level government wastes committed to be moved to Yucca Mountain by specified dates? And (b) can the $13.6 billion invested at Yucca Mountain be salvaged if the NRC approves the license submittal and the opposition relents after contentious hearings? Or will it take contingent actions, or, a switch to a partial closed fuel cycle with its reduced risks and earlier timing of their peak risk value? Only time will tell if any of these alternates will be acceptable but, they all reinforce the need for interim storage for commercial SNF.If the decision is to go to a new repository one wonders whether the time has not come to change the safety evaluation process for geological repositories by characterizing two to three sites and subjecting them to an arbitrary release of the fission products in the HLW to be stored and considering the forms of some of the HLW to reduce their peak risks. It would allow the proper choice to be made among the selected sites and to have a basis for convincing the local committee to accept the repository location. It may even decide whether the CONFU fuel assembly [MIT, 2006. Implications of alternative strategies for transition to sustainable fuel cycles. Nucl. Sci. Eng., 154 (September)] for pressurized water reactors can be accommodated in a once-through fuel cycle as suggested by Levy [Levy, S., 2008. Yucca backup plan. Nucl. Eng. Int., 24–28]. A similar configuration is possible in boiling water reactors.     

10-MWt固态钍基熔盐堆乏燃料贮存系统临界安全影响分析

10-MWt固态钍基熔盐堆(Thorium-based Molten Salt Reactor-Solid Fuel,TMSR-SF)使用TRISO(Tri-structural isotropic)颗粒燃料元件,并采用熔融氟盐作为一回路冷却剂,附着在燃料元件上的熔盐有可能影响系统反应性。因此,需要分析在燃料元件的贮存过程中熔盐附着燃料元件对贮存临界安全的影响。使用SCALE6.1的TRITON(Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)模块对TMSR-SF堆芯建模并进行燃耗计算,使用MCNP对乏燃料贮存系统进行临界计算。分别考虑熔盐浸渗球形燃料元件和熔盐包覆在球形燃料元件表面两种典型情况下,熔盐附着对贮存系统反应性的影响。针对乏燃料贮存系统,以浸渗最大量,即熔盐体积是石墨体积的13.9%为前提,临界计算结果表明,熔盐浸渗入石墨基体贮存系统的反应性比熔盐包覆在球形燃料元件表面的贮存系统的反应性要大5%;与没有熔盐附着的情况相比,有熔盐附着的情况下贮存系统反应性要大15%。对乏燃料贮存系统的临界安全分析可知,两种典型的熔盐附着模型对贮存系统的反应性存在一定的影响,但无论是熔盐浸渗还是包覆,贮存系统仍处于次临界,意味着贮存系统在正常工况下是安全的。     

Topics of research and development on concrete cask storage of spent nuclear fuel

This paper addresses topics of research and development (R&D) being challenged for realization of concrete cask storage of spent nuclear fuel in Japan. Comparison between metal cask storage and concrete cask storage is addressed. Background of these R&D and current status of technology on spent fuel storage are described. Need and design concepts of concrete cask storage technology, tests and evaluation of integrity of spent fuel, materials, concrete casks under normal and accident conditions, monitoring technology, etc. are systematically arranged and introduced. Topical problems of these R&D are described.