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.     

Analysis of criticality in shipment and storage of fuel at a nuclear power plant with a VVÉR reactor

The substantiation of nuclear safety during shipment and storage of fresh and spent fuel at nuclear power plants with VVéR reactors is examined in the light of the more stringent nuclear safety rules. Possible technical measures for satisfying the safety criterion are examined, for example, the concept of subcritical fresh fuel. An example of the estimation of the probability of the formation of a critical mass as result of fuel assemblies falling randomly out of a container is presented. Certain characteristic features of the calculation of the neutron-physical characteristics of fuel in a cooling pond are presented, for example, the nonconservative nature of a separate analysis in the infinite approximation. 4 figures, 5 references. OKB “Gidropress”. Translated from Atomnaya éneriya, Vol. 87, No. 1, pp. 11–16, July, 1999.     

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.     

Preparation and removal for reprocessing of spent VVR-2 and or nuclear fuel from the Russian Science Center Kurchatov Institute

The details of the preparation and removal of spent nuclear fuel from the Institute’s VVR-2 and OR research reactors for chemical reprocessing are presented. The spent fuel is represented by fuel assemblies which have different shapes and contain EK-10 fuel elements with similar construction and UO₂–Mg 10% enrichment kernels or S-36 fuel elements with U–Al alloy kernels with 36% enrichment. The storage conditions for the spent fuel are described. The details of the procedures developed to identify fuel assemblies by type of fuel elements are presented. The choice of the TUK-19 shipment container for loading and transporting spent fuel for reprocessing is validated. The details of the loading of spent fuel assemblies into TUK-19 are described; these operations are performed by workers under a protective layer of water in a handling room specially designed for such purposes. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 201–209, April, 2009.     

Handling spent nuclear fuel at a temporary storage site in Guba Andreev

Spent nuclear fuel has been stored in dry-storage units at a shore base of the naval fleet for 35–45 year. The total activity of the spent nuclear fuel is 170 PBq. This article presents data which characterize the state of the fuel (from normal to defective), the radiation conditions, and information on the individual and collective irradiation dose to workers. The results of an inventory check of the cells and jackets which contain fuel assemblies are presented. The corrosion processes are described and ideas for handling the spent fuel at the RT-1 plant of the Mayak Industrial Association, including handling fuel assemblies and jackets in cases, are described. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 56–61, July, 2006.     

Use of low neutron leakage fuel loads to decrease the radiation load on a VVÉR-1000 vessel

It is shown on the basis of data obtained at Ukrainian nuclear power plants that fuel loads with low neutron leakage can be used effectively to decrease the radiation load on the reactor vessel. The characteristics of 104 fuel loads and the results of a determination of the radiation load on the vessel are analyzed to develop a criterion according to which a VVéR-1000 fuel load can be classified as a load with low neutron leakage. It is shown that the following condition can be chosen as such a criterion: the run-averaged relative power release in all protruding fuel assemblies must be less than 0.57. Different variants of the arrangement of the VVéR-1000 core are examined and analyzed. It is shown that placing burned-out fuel assemblies along the periphery of the core and decreasing the number of neutrons leaving the core do not always result in a lower neutron load on the reactor vessel. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 93–97, August, 2006.     

Safety of the storage of spent submarine reactor compartments with no nuclear fuel on or under water

This article is a continuation of the radiation safety analysis of storage of spent submarine compartments with no nuclear fuel in an inlet. The first results were, published in Atomnaya énergiya,85, No. 3, 233–238 (1998). In contrast to the previous analysis, the danger of leakage of radioactive substances not only from shielding materials but also from a PWR type reactor with no fuel is examined. Marine storage of spent compartments in a sealed container is shown to be safe under conditions such that water flows past the container at a rate of 10⁷ m³/yr. 1 figure, 4 tables, 7 references. Russian Science Center “Kurchatov Institute” Translated from Atomnaya énergiya, Vol. 86, No. 3, 225–232, March, 1999.     

Reactor with a fast-resonance neutron spectrum, cooled by supercritical-pressure water with a bidirectional coolant flow scheme

A scheme for circulating coolant and cooling the core that has advantages over the designs of similar nuclear power systems is proposed for light-water reactors with supercritical coolant parameters and a fast-resonance neutron spectrum. A negative void coefficient of reactivity is obtained for the entire run of a fuel assembly without building a blanket. A more uniform distribution of the energy release over the core volume is achieved without using complicated fuel-enrichment schemes. The nonuniformity of the coolant temperature distribution at the core exit is decreased. The fuel assemblies operate with a much lower temperature drop over the core height. The core has a small reactivity excess on burnup and a BR of about 1, for which the most difficult operating regimes (flooding with cold water) can be handled with standard means (placement of absorbing organs of the safety and control system in ∼2/3 of the fuel assemblies). __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 349–356, May, 2006.     

Method for identifying 17-place casings with spent fuel assemblies of AMB reactors at the Beloyarskaya nuclear power plant

A method is proposed for identifying a cylindrical case with a jacket containing 17 spent fuel assemblies from the AMB reactors at the Beloyarskaya nuclear power plant on the basis of measurements of the radiation characteristics in lateral surface of the jacket opposite the fuel assemblies in the outer and inner rows. Computational validation using the PRIZMA and MCNP computer codes is given for the method. It is shown that the collection of signals from detectors is specific to each jacket, and this makes it possible to use it as an identifying indicator. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 363–367, June, 2007.     

Nuclear safety of spent fuel pool storage at nuclear power plants with VVER-1000: physical principles and structural solutions

The physical principles and structural solutions concerning nuclear safety in handling nuclear fuel at nuclear power plants with VVER-1000 are examined. It is shown that a structure of storage racks in the holding pools based on the spacing of hexahedral boron steel pipes, the structure acting as a combined neutron trap, will increase the pool capacity as compared with the loosely packed pools and will make the handling of nuclear fuel nuclear safe under normal operating conditions and during accidents. In the future, this construction will make it possible to use fuel assemblies with high uranium content and fuel enrichment above 5%.     

Basic principles for ensuring nuclear safety in designs of transportable packaged assemblies

Conclusions Any design of transportable packaged assemblies intended for the conveyance of spent nuclear fuel must meet the requirements of PBYa-06-88-77 and must give representative evidence of nuclear safety. The proposed classification of damage established a suitable procedure for calculating the critical parameters of the system when carrying out the preliminary analysis of the nuclear safety of packaged assemblies at different stages of their design. After the preliminary analysis of the nuclear safety of the TK-6 and TK-10 packaged units, the acceptable significant and dangerous damage has been established, which leads to displacement of the fuel within the package, which creates a real possibility for the technical assurance of nuclear safety.Translated from Atomnaya Énergiya, Vol. 49, No. 4, pp. 216–218, October, 1980.     

Optimal fuel utilization during decommissioning of nuclear power plants with RBMK reactors

When RBMK reactors are decommissioned successively in the same nuclear power plant, part of the fuel of the stopped reactors can be transferred to other units and additionally burned in continuing operations. The problem of minimizing the consumption of fresh fuel by optimal distribution of the additionally burned fuel over the reactors is examined. The limitations on the refueling rates, the holding time of fuel assemblies prior to transfer, the service life of fuel assemblies, and certain characteristics of reactors are taken into account. It is shown that the reuse of fuel in other units permits saving from one to almost two thousand fresh fuel assemblies and that the effect of optimizing the additional burn regime can reach several hundreds of saved fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 284–290, May, 2007.     

Design and thermal modeling of radially configured nuclear waste packages

A series of radial design configurations for packaging nuclear wastes are described. These radial arrangements for used nuclear fuel assemblies in containers are effective techniques for packaging significantly more radioactive waste in the available internal container volume. The radial package designs can be applied to packaging the nuclear waste for permanent storage at the Yucca Mountain (YM) repository. The radially configured containers will have high degree of structural strength and will be efficient in transferring heat from the waste form to the package surface due to the minimization of internal gaps. Radial configurations are reported for packaging the Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) used fuel assemblies. These configurations can be varied for co-packaging the colder, i.e. vitrified high level waste (HLW), canisters. Details of the geometry and the materials selected are discussed. Thermal analysis of the radial designs was conducted which confirm the feasibility of the designs demonstrating that no over-heating occurs in the contained nuclear waste in spite of the significantly extra amount of waste. The larger amount of packaged waste per container coupled with efficient heat transfer characteristics of these designs favor hotter and drier conditions for container surfaces in the YM emplacement drifts.     

核燃料组件无损检测探测系统设计

燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能X射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。     

Application of Fire-Retardant Treatment to the Wood in Type A Unirradiated Nuclear Fuel Outer Containers

Abstract

Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fireretardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented.     

Detection System for the Fuel Assembly Nondestructive Testing

Fuel assemblies are the central components of a reactor. The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature, high pressure and intense neutron radiation field, which threats the safety of the reactor. To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up, high-energy X-ray non-destructive testing is an effective technical means. Irradiated nuclear fuel assembly has a strong radioactivity, it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assembly during detection system designing phase. Through modeling, estimating and optimization, the optimal size of the detector unit is obtained and the collimator design is optimized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.     

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 35–44, January, 2005.     

Nondestructive assay of spent nuclear fuel with gamma-ray spectroscopy

An important issue in nuclear safeguards is to verify operator-declared data of spent nuclear fuel. Various techniques have therefore been assigned for this purpose. A nondestructive approach is to measure the gamma radiation from spent nuclear fuel assemblies. Using this technique, parameters such as burnup and cooling time can be calculated or verified.     

Spectrometric method of evaluating spent nuclear fuel characteristics

A spectrometric method of identifying spent fuel assemblies according to the type of fuel elements present in them is described. The method is based on the results of spectrometric measurements and subsequent analysis of the radiation from fission products and the characteristic radiation from uranium in the irradiated fuel. The fuel assemblies used in the VVR-2 and OR research reactors contained fuel elements of a different type, differing by the initial quantity of uranium contained in them. To prepare the spent fuel assemblies for shipment to a reprocessing facility after long-time storage in cool-down pools, the assemblies must be sorted according to the type of fuel elements present in them. The method developed for identifying the types of fuel elements in the irradiated fuel is based on the dependence of the intensity of the characteristic radiation from uranium on the uranium content in a fuel element. The degree of excitation of the characteristic radiation of uranium also depends on the intensity of the radiation from fission products, which is monitored during the spectrometric measurements performed on the irradiated fuel; ultimately, this makes it possible to sort the spent fuel assemblies.     

Application of gamma-neutron ionization chambers for monitoring the distribution of the neutron flux density,energy release,and energy production in RBMK-1000

A method developed for performing direct measurements of three-dimensional distributions of energy release and energy production in RBMK fuel assemblies is described. The method is based on performing measurements with a gamma-neutron chamber and comparing the neutron and gamma signals. The results of the measurements of the neutron flux density, energy release, and energy production are compared with the values obtained with the Prizma-M program of the Skala-micro information-measurement system. It is confirmed experimentally that the Prizma-M system can be used to monitor the distribution of not only the neutron flux density and energy release of fuel assemblies but also the energy production of off-loaded fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 182–186, September, 2007.