Ways of saving steel in reactor construction

The basic methods for saving steel in reactor construction are considered in this article. It is shown that, in stationary nuclear reactors, the basic method for saving steel is its replacement with concrete. The article provides data indicating that the mechanical strength and the radiation resistance of concretes are not being utilized to the fullest extent at the present time, and that capillary cracks, which appear in concrete, do not affect its protective properties. The problem of the application of protective explosion-proof shells is considered. The project of the Soviet Boiling Reactor is described as an example proving the efficiency of using a protective chamber (instead of a protective shell) which is designed for excess internal pressure.     

Tandem mirror reactor with thermal barriers

A new invention — the thermal barrier — promises to improve the tandem mirror fusion reactor. The thermal barrier consists of a region of reduced magnetic field strength, plasma density, and plasma potential between each end plug and the central cell of a tandem mirror. The depressed plasma potential serves to thermally insulate the plug electrons from the central cell electrons. With barriers and auxiliary electron heating in the plugs, the central cell confining potential can be generated with a lower plug plasma density, magnetic field strength, and beam injection energy than for the case without barriers. This paper summarizes the status of the rapidly evolving physics knowledge concerning tandem mirrors with thermal barriers, describes end plug components typical for tandem mirror reactors — yin-yang magnets, neutral beams, and ECRH heating systems, and discusses central cell design.     

Some new viewpoints in reactor noise analysis

It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model‘s phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.     

Some problems of pulsed reactor dynamics

Selected problems concerning the dynamics of a fast pulsed reactor are considered; firstly, the method is discussed of pulse generation in a fast reactor by means of insertion of a rod with the presence in the reactor of a strong neutron source; an equation is derived for the optimum velocity of the rod; solutions are given of dynamical equations relating to the case of movement of the rod; secondly, the effect is investigated of a large internal cavity on the energy release following a pulse and on the reactivity thermal quenching factor for a fast pulsed reactor in spherical geometry.Translated from Atomnaya Énergiya, Vol. 16, No. 4, pp. 309–314, April, 1964     

Space-energy distribution of thermal neutrons in a homogeneous reactor

The space-energy distribution of thermal neutrons in a heterogeneous reactor is determined in a heavy gas model approximation. The moderator is assumed to be nonabsorbing and the fuel sings are assumed to be nonmoderating to neutrons. In order to describe the energy dependence of the solution, adjoint functions are included in addition to the Laguerre polynomials normally used, and which is shown to be very suitable. Possible methods of refining the results are discussed.Translated from Atomnaya Énergiya, Vol. 17, No. 3, pp. 193–198, September, 1964     

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.     

Some nonlinear problems of nuclear reactor theory

In reactor criticality calculations it is necessary to take into account the relationship between the tamperature and volume of the reactor and the neutron flux. It is frequently assumed for this, that k_eff is a function of the reactor power. The method proposed here is an approximate computation of this relationship by means of some functional of the flux (this functional can be expressed by the power averaged with respect to the reactor temperature, etc.), so that its calculation by means of k_eff is a special case of the proposed method. Results are given of calculations for a system of nonlinear equations which describe the neutron transport in one-group diffusion approximation in the plane of the reactor and the heat transfer by thermal conductivity. The results are analyzed with the object of comparing the precise and approximate solutions.Translated from Atomnaya Énergiya, Vol. 16, No. 4, pp. 304–309, April, 1964     

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.     

Static strength analysis of CANDU-6 reactor fuel bundle

A static analysis, finite-element (FE) model was developed to simulate out-reactor fuel–string strength tests with use of the well-known, structural analysis computer code ABAQUS. The FE model takes into account the deflection of fuel elements, and stress and displacement in endplates subjected to hydraulic drag loads. It was adapted to the strength tests performed for CANFLEX 43-element bundles and the existing 37-element bundles. The FE model was found to be in good agreement with experiment results. With use of the FE model, the static behavior of the fuel bundle string, such as load transfer between ring elements, endplate rib effects, hydraulic drag load incurring plastic deformation in fuel string and hydraulic flow rate effects were investigated.     

Investigation of the thermal inertia and thermal stresses of heat-transferring elements in a solid-coolant nuclear reactor

Computational studies of specific questions arising when solids are used in nuclear reactors as the cooling medium are described. The concept of using approximately 1 mm in diameter spherical heat-carrying elements made of pyrolytic carbon coated graphite as a medium for transmitting heat from a fuel element to a steam generator is examined. A computational analysis of the internal stresses arising in 1–10 mm in diameter spherical elements transferring heat and the temperature lag of the heat-carrying elements relative to the temperature of the medium under cyclic heating and cooling is performed. The results of experimental studies were used to determine the boundary conditions of the problem. Supplying heat uniformly over the surface and through a finite number of contact points, which is characteristic for a fill consisting of spherical particles, was modeled. The transmission of heat through a finite number of contacts results in a complicated stress state of the heat-transmitting elements and a higher thermal inertia. It is shown that the internal stresses are weak in small heat-carrying particles, but when the diameter is increased to 10 mm the stresses from thermal cycling become comparable to the ultimate strength.     

Study on reactor building structure using ultrahigh strength materials

This study was promoted to realize the optimal nuclear reactor building structure of the future. As the first step, the study regarding ultrahigh strength reinforced-concrete (RC) shear walls was positively verified. The tests conducted were relevant to ultrahigh strength concrete material tests, pure shear tests of RC flat panels, bending shear tests and simulation analyses of RC shear walls, S-joint tests and mixed-structure tests. From the results of this study, it was verified that mixed structures using ultrahigh strength material can be realized.     

热中子研究堆薄板型控制棒价值计算

对通过采用能量相关的内边界条件,在扩散理论的范围内计算薄板型热中子研究堆控制棒价值进行了理论分析,成功地计算了日本原子能研究所JRR-3M的控制棒价值,并与日本的计算结果进行了对比。