Obtaining molybdenum-99 in the IRT-T research reactor using resonance neutrons

The yield of ⁹⁹Mo from the ⁹⁸Mo(n,γ)⁹⁹Mo reaction significantly depends of the energy spectrum of the neutron flux. It is well known that the cross-section for this reaction is about 130 mb, whereas the resonance integral of the reaction is 6.9 b. The aim of this work was to investigate the conditions that let to increase ⁹⁹Mo yield from the targets with natural and enriched isotope composition under irradiation by resonance neutrons at the IRT-T research reactor.The calculations of integrated cross-sections of all Mo isotopes in the region of the ⁹⁸Mo resonances showed that screening in the target with natural isotope composition by other isotopes is relatively small. So the ⁹⁸Mo in the natural mixture can be activated by resonance neutrons approximately in the same manner as pure ⁹⁸Mo.Experimental measurements of the ⁹⁸Mo(n,γ) effective cross-section using the MoO₃ sample with natural and enriched composition in the reactor channels with the beryllium moderator with the thickness of 20 up to 90 mm showed that the effective cross-sections in these channels reach the value of 700 mb. The contribution of the epithermal neutrons into the ⁹⁸Mo activity was 68% for the enriched targets and 78% for natural molybdenum, respectively.At that channel it is possible to produce ⁹⁹Mo with specific activity up to 3.4 Cu/g with samples of natural isotope composition and up to 15 Cu/g with enriched samples on the base of reactors with neutron flux of (1.7 × 10¹⁴ n/(cm² s)). Such ⁹⁹Mo specific activity is enough not only to realize extraction technologies production of ^99mTc, but to manufacture sorption generators of ^99mTc without wastes.     

铀水栅零功率反应堆的功率刻度

反应堆功率的测量,在堆功率高时一般用热工方法,功率低时,可用各种堆物理方法,如中子源引进法、中子统计法和全堆总裂变率法。 中子源引进法误差较大,中子统计法需知探测器在堆内的效率和堆的β_(aff)值,此二者都较难测量。全堆总裂变率法是由测量堆的总裂变率来求得堆功率,它可避免前面两种方法的缺点,但需依赖裂变率相对分布的     

Status of the small modular fluidized bed light water nuclear reactor concept

The state of the art of a small modular reactor concept with a suspended core is presented. The reactor design is based on a fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident conditions. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another option is to modify the shape of the reactor core to produce a non-fluctuating bed and, consequently, guarantee the dynamic stability of the reactor. The mixing of tantalum in the fuel is also proposed as an additional inhibition to the power excursion. The spent fuel pellets may not be considered nuclear waste, since they are of a shape and size that can easily be used as a source of radiation for food irradiation and industrial applications. The reactor can easily operate as a plutonium burner or can operate with a thorium fuel cycle.     

实验堆水平实验孔道闸门启闭驱动装置设计

高通量实验反应堆有多种水平实验孔道，它的垂直闸门的开启与关闭需要驱动机构来实现。选择以电动-机械传动的方式，设计了具有l0吨承载能力的电动千斤顶。这种传动方式的千斤顶具有结构合理．传动灵活、技术上安全可靠的特点．符合高通量实验堆水平孔道垂直闸门的设计要求。     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well.     

Model of the Destruction of the Reactor in the No. 4 Unit of the Chernobyl Nuclear Power Plant

The authors, who participated in 1986–1995 in the investigations of the destroyed power-generating unit of the Chernobyl nuclear power plant, have made an attempt to reconstruct, on the basis of an analysis of the results of investigations, the processes leading to the destruction of the reactor system. It is shown how the ideas concerning the accident processes developed as data on the postaccident state of the structures of the reactor system were obtained and analyzed. A model of the destruction of the reactor system is proposed. This model substantiates that fact that the core was lifted by 30 m from the reactor shaft and exploded above the central hall.     

氟盐冷却球床堆与HTR-10和MSRE的定量相似性分析

氟盐冷却球床堆是当前国际上一种新的研究堆型,尚无已经建造完成的反应堆,因此,选择相似且具有运行经验的反应堆作为基准题有助于堆芯核设计软件适用性分析。利用国际上常采用的相似性分析软件,可对熔盐实验堆(Molten Salt Reactor Experiment,MSRE)及10 MW高温气冷堆(10 MW high-temperature gas-cooled test reactor,HTR-10)与氟盐冷却球床堆的相似性进行分析,定量判断它们作为基准题的合理性。分析结果表明,MSRE和氟盐冷却球床堆的能谱峰位能量接近且堆内元素种类相近,二者相似程度较高;常温临界HTR-10和氟盐冷却球床堆冷却剂不同,且能谱峰位能量差异较大,二者相似程度较低。因此,MSRE是氟盐冷却球床堆中子物理设计软件较理想的基准题。     

Study on the passive safe technology for the prevention of air ingress during the primary-pipe rupture accident of HTGR

A primary-pipe rupture accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This study is to investigate the air ingress phenomena and to develop the passive safe technology for the prevention of air ingress and of graphite corrosion. This paper describes the method for the prevention of air ingress into the reactor during the primary-pipe rupture accident. It is found that a safe cooling rate of the reactor core exists for the prevention of air ingress. The experimental results show that the natural circulation flow of air during the accident can be controlled by the method of helium gas injection into the reactor pressure vessel.     

Control of the power of a fast reactor in the self-regulation regime and reactor startup

An algorithm for controlling the power of a nuclear reactor operating in the self-regulation regime and reactor startup is proposed. The algorithm is based on investigations of a dynamical model of the reactor. It makes it possible to start up the reactor and maintain it quite accurately at a prescribed power level. However, an additional safety investigation is needed.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 18–24, January, 2005.     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

The application of mechanical and thermal cutting tools for the dismantling of activated internals of the reactor pressure vessels in the Versuchsatomkraftwerk, Kahl and the Gundremmingen Unit A

The Gundremmingen Unit A plant (KRB A) and the Versuchsatomkraftwerk (VAK) plant represent the first generation of nuclear reactors in Germany. The 250 MW_e reactor KRB A was the first commercial reactor in Germany and the 16 MW_e reactor VAK was the pilot nuclear power plant, which had to serve mainly scientific purposes. KRB A is under dismantling since 1983, VAK since 1988. Although they are both of the boiling water type, they are rather different to each other, referring to their size and construction. The actual work is the dismantling of high contaminated components inside the reactor buildings and the underwater cutting of activated internals of the reactor pressure vessels. Several cutting techniques have been developed, tested and applied to respective dismantling tasks in the meantime. The experiences made in both projects are not limited to dismantling work only, but also include know-how on effective decontamination and scrap recycling.     

Preliminary Design and Analysis of the DC Reactor for the ITER Converter

The DC reactor is an important piece of equipment for restraining loop and ripple currents in the international thermonuclear experimental reactor （ITER） converter power supply system. As the reactor is operated at a steady state of 27.5 kA and needs to withstand a peak current of 175 kA, so the design of the DC reactor used in the ITER converter power supply system is necessary. A new water-cooling dry-type air-core reactor is designed in this work. The detailed structural parameters are calculated by theoretical formulas, and then the structure is optimized by electromagnetic simulation with ANSYS. Finally, thermal and dynamic stability analyses are performed to verify the temperature and stress at a rated current of 27.5 kA and pulsed current of 175 kA. The analysis results show that the temperature and stress meet the requirements of the ITER converter power supply system.     

Ion beam analysis of materials in the PBMR reactor

South Africa is developing a new type of high temperature nuclear reactor, the so-called pebble bed modular reactor (PBMR). The planned reactor outlet temperature of this gas-cooled reactor is approximately 900 °C. This high temperature places some severe restrictions on materials, which can be used. The name of the reactor is derived from the form of the fuel elements, which are in the form of pebbles, each with a diameter of 60 mm. Each pebble is composed of several thousands of coated fuel particles. The coated particle consists of a nucleus of UO₂ surrounded by several layers of different carbons and SiC. The diameter of the fuel particles is 0.92 mm. A brief review will be given of the advantages of this nuclear reactor, of the materials in the fuel elements and their analysis using ion beam techniques.     

Physics features of the HTR for process heat

The concept of the pebble bed reactor and of the fuel elements allow flexible adaption to various requirements of the reactor application. Beyond electricity it opens the use of the nuclear power in the market of thermal and chemical energy. The inert and equalizing response of the reactor in accidental situations makes possible the location of the reactor in the vicinity of the factories and living areas which use the power. Major flexibility is given when composing a larger power plant out of smaller reactor units in a modular mode. The application of the thorium fuel cycle with recyling of the bred uranium allows careful protection of the natural uranium resources. These characteristics make the pebble bed reactor an attractive concept for the advanced utilization of the nuclear power in the world.     

Optimization of the reflector design of the Syrian MNSR

The possibility of optimizing the dimensions and material of the miniature neutron source reactor of Syria was investigated based on a 3-D model of the reactor comprehensive of all reactor components. It was shown that the reflector can be optimized in terms of materials to have 10.00–11.50 mk of initial excess reactivity available at the start-up of the reactor if the side reflector material were substituted by the bottom reflector material. Otherwise if the initial excess reactivity of ～4.00 mk were to be maintained the saved material would be about 15.000 kg.     

Mechanical Desktop软件在中国先进研究堆堆本体设计中的应用

研究使用MechanicalDesktop(MDT)软件对中国先进研究堆 (CARR)堆本体主要部件进行三维参数化建模 ,并通过尺寸及相关位置数值的变量驱动进行CARR堆本体初步设计及修改。三维参数化设计方法的应用大大提高了CARR堆本体的设计效率 ,缩短了设计周期 ,为高质量如期完成CARR堆本体主要部件的设计奠定基础