Neutron-physical and thermohydraulic calculations of VVR-SM with high-and low-enrichment uranium fuel assemblies

The VVR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences of Uzbekistan is being converted from fuel assemblies with high-enrichment uranium (36% ²³⁵U) to fuel assemblies with low-enrichment uranium (19.7% ²³⁵U). During the conversion process consisting of nine cycles, the IRT-3M fuel assemblies with high-enrichment uranium, which are removed at the end of each cycle, will be replaced with IRT-4M fuel assemblies with low-enrichment uranium. This will require increasing the core size up to 20 fuel assemblies and increasing the power of the reactor to 11 MW. The methods used for and the results of neutron-physical calculations (burnup, power distribution, subcriticality), thermohydraulic analysis, and calculations of the kinetic parameters of a stable state are described for a core with high-enrichment uranium, a mixed core, and the first full core with low-enrichment uranium. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 269–273, May, 2008.     

Service life extension of vvr-sm fuel assemblies

Some operating properties of VVR-SM are discussed. It is shown that the thermal load on fuel elements can be greatly decreased in a definite powering up regime and, in consequence, the fuel-assembly service life can be extended and the radiation conditions can be improved.     

Operation of VVR-SM with 36% Enrichment Uranium-Dioxide Fuel

VVR-SM after conversion to IRT-3M fuel assemblies with 36% fuel enrichment is operating at 10 MW with a core load of 18 fuel assemblies. The safety of the operation is ensured by neutrons and thermohydraulic calculations of the operating regimes of the core after each reloading of fuel assemblies. The experimental channels of nine fuel assemblies with flux density of neutrons with energy >0.821 MeV up to 1.1⋅10¹⁴ sec⁻¹⋅cm⁻² are used for producing ³²P and ³³P. The flux density of thermal neutrons in the channels of the beryllium reflector reaches 1.6⋅10¹⁴ sec⁻¹⋅cm⁻². Nine of the horizontal channels and the channel of the thermal column are used for fundamental and applied studies and for neutron-activation analysis. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 147–152, August 2005.     

An Emergency Response Plan for Transport

Abstract

Transnucl6aire is involved in road and rail transport of nuclear fuel cycle materials. To comply with IAEA recommendations, Transnucl6aire has to master methods of emergency response in the event of a transport accident. Considering the utmost severe situations, Transnucl6aire has studied several cases and focused especially on an accident involving a heavy cask. In France, the sub-prefect of each department is in charge of the organisation of the emergency teams. The sub-prefect may request Transnucl6aire to supply experts, organisation, equipment and technical support. The Transnucleaire Emergency Response Plan covers all possible scenarios of land transport accidents and relies on: (i) an organisation ready for emergency situations, (ii) equipment dedicated to intervention, and (iii) training of its own experts and specialised companies.     

Public Safety Aspects of Nuclear Power Plant Licensing

Various aspects of the siting, design, construction and operation of nuclear power plants to which the Atomic Energy Commission gives attention in its licensing activities are discussed, together with the reasons why they are considered important to public safety. This includes the measures taken to assure the integrity of the nuclear fuel and primary coolant system under all operating conditions; to provide adequate means of coping with emergency situations; and to limit the consequences of accidents to levels which do not endanger the health and safety of the public.     

Modelling of PWR cladding corrosion in mixed core situations

In this paper a thermal-hydraulic model for cladding corrosion recently developed in ABB Atom and used in the code is presented. The features of the model are a subchannel geometry which consists of a 3 × 3 matrix of rods, and modelling of coolant cross-flow and coolant enthalpy mixing. The thermal-hydraulic model is benchmarked against the code, which is a 3D code for analysing the thermalhydraulics of a reactor core. In addition, results of model calculations are compared with corrosion data obtained in mixed core situations, i.e. situations where the fuel assemblies in the core have different designs (e.g. different grid and nozzle designs). Fuel assembly components in assemblies of different designs usually have unequal flow resistances. These differences result in transverse pressure gradients, which in turn increase the lateral flow velocity and thus affect the cociant mass flow rate distribution. Two different situations where this type of mismatch between fuel assemblies in the Ringhals 3 core have occurred are studied in this paper. In the first case a reload batch of fuel assemblies, with Zircaloy mixing vane grids, inserted in a core where the resident fuel assemblies have Inconel mixing vane grids is considered. In the second case cladding tubes from the same manufacturing lot that have been irradiated for the same period of time but have been situated in fuel assemblies with Zircaloy mixing vane grids of different designs are considered. The results manifest the capability of the code to model the effects of flow resistance on cladding corrosion.     

Results of the verification period 1985–1990 of the German-Swiss high converting reactor PWR-C1

A high converting reactor (HCR), the PWR-C1, is described, which was developed on the basis of the commercially proven pressurized water reactor (PWR) technology by Siemens AG Power Generation Group in cooperation with the Karlsruhe Nuclear Research Center (KfK), the Paul Scherrer Institute at Würenlingen (PSI) and the Technical University of Braunschweig (TUBS).The essential features of the concept in comparison to the conventional Konvoi-reactor core are a triangular fuel pin lattice with a tighter pitch and a higher density of control assemblies. The main goal was a better fuel utilization, which was achieved mainly by hardening of the neutron energy spectrum. Because of the lower moderation ratio due to the tighter pitch, the neutron spectrum is shifted into the epithermal energy range, whereby the conversion rate is augmented, from 0.35 for a Konvoi-reactor core to 0.75 for a PWR-C1. The PWR-C1 concept is the result of extensive investigations of tight and very tight fuel pin lattices, concerning neutron physics, thermohydraulics, emergency cooling and mechanical design. For establishment and improvement of codes several experiments were performed in the mentioned areas; (1) the PROTEUS experiments for physics investigations at PSI Würenlingcon, (2) the CHF experiments at Siemens Karlstein and KfK, and (3) the NEPTUN and FLORESTAN flooding experiments at PSI and KfK. All these experiments led to the validation or improvement of codes and/or the verification of calculational methods. Also methods used in conventional design activities could be improved.The investigations in mechanical design too led to results, which partly are of great use for conventional design work.     

Effect of the VVR-SM neutron spectrum on the radioactivity and color of natural topazes

The neutron flux density from 0.025 eV to 12 MeV has been measured experimentally in all channels of the VVR-SM core by the activation method using threshold monitors (Au, Ni, Fe. Ti, Mg, Y). Comparing with a calculation of the neutron flux density at different energy using the IRT-2D computer code showed agreement to within 5%. The distribution of the neutron fluxes and spectra in the core, which is of practical utility for radiation technologies, was obtained. A series of irradiations has been conducted and experimental dependences of the irradiation time on the channel position in the core as well as on the size of the stones for obtaining a standard light blue and dark blue color have been obtained. The irradiation conditions making it possible to lower the induced radioactivity of the minerals three-fold as a result of increasing the ratio of the fast to thermal neutron fluxes are found.     

Revisiting the concept of HTR wallpaper fuel

As early as the 1970s, attempts were made to reduce the peak fuel temperature by means of so-called “wallpaper fuel”, in which the fuel is arranged in a spherical shell within a pebble: By raising particle packing fraction, fuel kernels are condensed to the outer diameter of the fuel zone, leaving a central part of the pebble free of fuel. This modification prevents power generation in this central fuel-free zone and decreases temperature gradient across the pebble.Besides particle temperature reduction, the wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. To assess such improvements, calculations were performed using Monte Carlo neutron transport and depletion codes MCNP/MCB. Among others, investigations of conversion ratio, temperature coefficient of reactivity, spent fuel composition and neutron multiplication (for which a method to determine the six-factor formula was developed), were conducted.It is demonstrated that this fuel type impacts positively on the fuel cycle, reduces production of minor actinides (MA) and improves the safety-relevant parameters of the reactor. A comparison of these characteristics with PBMR-type fuel is presented: By comparison with PBMR fuel, the “wallpaper design” results in an effective neutron multiplication coefficient increase (by about 1750 pcm), which is combined with a decrease of between 4.6 and 17.5% in MA production. An improved neutron economy of the heterogeneous design enables enrichment of the “wallpaper type” of fuel to be reduced by more than 6%.The fuel changes suggested in this paper offer more versatility to the HTR concept: Conversion ratio can be decreased (leading to lower MA build-up and fuel reprocessing cost) or raised (leading to lower fuel consumption and fuel cost). Variations around this concept also enable higher reactivity, thus higher achievable burn-up, improving sustainability of HTRs.     

Reprocessing free nuclear fuel production via fusion fission hybrids

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th²³²–U²³³ conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO₂ matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U²³⁵ fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

An analytical solution to fuel-and-cladding model of the rewetting of a nuclear fuel rod

An exact solution of the quasi-steady two-dimensional conduction equation for the rewetting of a nuclear fuel rod in water reactor emergency core cooling is obtained for a fuel-and-cladding model. A method of solving non-separable differential equations is presented, which is used in the present analysis. The recently developed theorem of orthogonality of piecewise continuous eigenfunctions is also used to handle the composite rod in the present model. The present analysis reveals that the wet front velocity increases with the increase of the gap resistance between the fuel and the cladding, and approaches a limiting value, which is equal to the wet front velocity of the tube of cladding alone, as the gap resistance becomes infinite. For convenience in practical application, the results of the present analysis are correlated in simple expressions.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

The 2007 Recommendations of the International Commission on Radiological Protection. ICRP publication 103

These revised Recommendations for a System of Radiological Protection formally replace the Commission's previous, 1990, Recommendations; and update, consolidate, and develop the additional guidance on the control of exposure from radiation sources issued since 1990. Thus, the present Recommendations update the radiation and tissue weighting factors in the quantities equivalent and effective dose and update the radiation detriment, based on the latest available scientific information of the biology and physics of radiation exposure. They maintain the Commission's three fundamental principles of radiological protection, namely justification, optimisation, and the application of dose limits, clarifying how they apply to radiation sources delivering exposure and to individuals receiving exposure. The Recommendations evolve from the previous process-based protection approach using practices and interventions by moving to an approach based on the exposure situation. They recognise planned, emergency, and existing exposure situations, and apply the fundamental principles of justification and optimisation of protection to all of these situations. They maintain the Commission's current individual dose limits for effective dose and equivalent dose from all regulated sources in planned exposure situations. They reinforce the principle of optimisation of protection, which should be applicable in a similar way to all exposure situations, subject to the following restrictions on individual doses and risks; dose and risk constraints for planned exposure situations, and reference levels for emergency and existing exposure situations. The Recommendations also include an approach for developing a framework to demonstrate radiological protection of the environment.     

Improvement of the design of the series-produced fuel element of the VVÉR-440 reactor to increase burnup

Conclusion The fuel elements for the VVÉR-440, which were developed for a 4-yr run, operate satisfactorily to an average burnup of 59 MW-day/kg of uranium in the most-stressed fuel element with average burnup of the unloaded fuel of 40 MW-day/kg. Through the introduction of end beveling of the pellets, the mobility of the fuel column is increased. Furthermore, the presence of bevels makes possible a reduction of the number of chips and elimination of process crumbs, and the high mobility of the fuel column makes it possible, during outfitting of the fuel element, to eliminate the axial gaps between pellets, which are unacceptable for safety reasons. The increase in the initial helium gage pressure has made possible substantial improvement of the thermomechanical characteristics of the fuel elements and avoidance of high fuel temperatures and large gas release. When the conversion is made to a 4-yr fuel run, the number of fuel assemblies refueled each year decreases from 117 to 90 per power-generating unit, natural uranium consumption is reduced by 11–12% [2], zirconium consumption is lowered, and the effective capacity of burned-up fuel stores is increased.I. V. Kurchatov Institute of Atomic Energy. Mashinostroitel'nyi Zavod Production Association. All-Union Scientific Research Institute of Aviation Materials. Translated from Atomnya Énergiya, Vol. 72, No. 2, pp. 121–124, February, 1992.     

Thermal-hydraulic analysis of accidents leading to local coolant flow decrease in the main circulation circuit of RBMK-1500

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.     

实施ICRP建议的主要挑战:近期和未来的发展（英文）

自2007年国际放射防护委员会(ICRP)发布一般性建议以来,在解决将这些建议付诸实践所面临的挑战方面取得了一些进展。值得注意的是,为放射防护管理而提出三类照射情况后(即,计划照射、现存照射及应急照射),我们进一步研究了它们在几个领域的应用,如涉及天然放射性物质的工业领域以及核事故发生后的应急与恢复领域。本文着重介绍了最近一些出版物提出的以及持续开展的工作中所涉及的主要问题,重点围绕与放射防护体系伦理基础有关的研究进展、放射防护体系在管理现存照射情况中的应用以及环境放射防护与ICRP防护体系的融合。最后,本文介绍了2022年ICRP成立的与放射防护体系应用相关的新工作组的工作目标。