Neutronics reactor core analysis

A critical assembly is a split bed on which fissionable material used to mock up up a separated reactor core that is stacked half on each half. One half is on roller guides so that the two halves may be quickly pulled apart if the neutron multiplication gets too high. Use the Preliminary Hazards Analysis method described in section 3,2.1 to identify the possible accidents that may occur and the qualitative probabilities and consequences. List the initiators in a matrix to systematically investigate the whole process. Don t forget human error. 

The amount of unreacted target element that eluted was determined by measuring its radioactivity directly in the case of actinides, and by activation analysis in the case of lanthanides. The distribution of the radioactive neutron capture product was determined by counting both the eluate and the eluted zeolite. All irradiations were done in the Oak Ridge Research reactor in a pneumatic tube facility with a thermal neutron flux of about 4 X 1013 neutrons cm-2 sec-1 or, for a few long irradiations, in a tube adjacent to the reactor core at the fluxes stated in Table VI. 

Core analysis includes determination of reactor core neutron flux profiles, determination of the conditions for reactor criticality, the effects of fuel burnup and fission product poisons on core flux and core temperature profiles based upon heat generation rate, coolant flow rate, and core inlet temperature. 

Diffusion theory still remains the main tool for fast reactor numerical analysis. Two reasons are mainly responsible for this (1) in the central part of the reactor, the reactor core, the application of the diffusion approximation is satisfactory for most purposes, (2) the diffusion equation is relatively simple, its properties and the appropriate methods to solve it have long been well known. Although the diffusion equation was applied previously in the other field of theoretical physics, the numerical methods used to solve efficiently the neutron diffusion problems encountered in reactor physics had to be tailored to the properties of the parameters appearing in the equation and to the neutron distribution as a function of space, energy and time. 

The scope of studies and computer codes are summarized in Table 1.15. SRAC is a neutronic core calculation code developed by Japan Atomic Energy Agency (IAEA) [21]. FEMAXI-6 is a light water reactor fuel analysis code also developed by IAEA [31]. All other codes in the table were developed mostly by graduate students at the University of Tokyo during their thesis studies. Some codes did not have names. 

Preliminary reactor core concepts have been proposed at various institutions. Among them, the concept with mixed neutron spectrum proposed by SJTU [95,96] has achieved the special attention of Chinese researchers. The mixed spectrum SCWR core combines the merits of both thermal and fast spectra as far as possible. The basic idea is to divide the reactor core into two zones with different neutron spectra. In the outer zone, the neutron energy spectrum is similar to that of PWRs. To assess the performance of the reactor core, a coupled neutron-physics and thermal-hydraulics analysis was conducted [96]. 

Neutron-capture prompt-gamma ray activation analysis (PGAA) is a recent addition to the nuclear analytical arsenal. In this technique the instantaneous gamma ray emission from a sample is measured as it is irradiated in a flux of reactor neutrons (33,3, 35). Because the sample must be several meters from either the core of the reactor or (less commonly) from the detector, the sensitivity of this technique is generally poorer than in conventional NAA. However, it is possible to measure small quantities of many elements which do not give radioactive neutron-capture products, notably 0.01 mg of H, 50 ng B, and 1 mg P in an electronics context. 

The use of activation methods for the analysis of biological material has been reviewed by Bowen ( 7). In Instrumental Neutron Activation Analysis, the dried sample is placed in the core of a nuclear reactor where it is bombarded with neutrons. Many of the elements present in the sample undergo nuclear reactions of which the most common are the (n, y) type. The products of these reactions are radioactive and decay with the emission of gamma photons of characteristic energy. 

A proper analysis of the tine dependent behavior of.a reactor operating on thermal neutrons must take into account the important effects on its criticality, reactivity, and stability which arise from such factors as fission i products of high thermal-neutron capture cross-section, depletion, temperature, average neutron lifetime in the reactor, flux level, and reactor period. As has been seen in the requirements placed on the.reactor, considerable excess reactivity must be built into the active core before start-up. The control rods must keep the reactivity below the critical value before and during start-up. 

Design Basis Accident analysis codes PREDIS and VENUS have been validated against European LOFA benchmark problems. Fuel subassembly worths at different radial positions in core during initial fuel loading were calculated. A study was made on the possibility of recriticality of molten fuel dropped in the core catcher from the core following an accident. Calculations of neutron irradiation dose for the reactor assembly out of core componets both in radial and axial locations were completed and indicated that the dose values are negligible ( ldpa). 

Analysis of the 90-element fuel storage container In use at Sandia Laboratories annular core pulse reactor (ACPR) was performed to characterize its subcritical multiplication properties under full and partial loading conditions. Neutron transport calculations were performed with the DTF-IV code (Sjf transport theory) and 16 energy-group cross sections. Experimental verification exists for some of the calculated arrays but the limited number of ftiel elements available (40) prevented a fill loading of the storage container. 

The thermal-hydraulic analysis is performed using special codes that receive the fission power spatial deposition from the reactor analysis pin power deconstruction and perform a heat transfer and fluid flow calculation to calculate the expected temperature profile throughout the core. Because neutron cross sections depend on temperature, this calculation must be iteratively linked into the previous six steps. In many cases, this linkage results in an expansion of the parametric expansions of the neutron cross sections (described as part of Step 6) to be a function of temperature and total assembly fission power. 

Also shown in the core map is a radial reflector assumed to consist of a 50 volume % HT9-50 volume % Pb mixture. The steel shroud surrounding the core is also represented by this region. A steel containing reflector is necessary to reduce the fast neutron fluence at the reactor vessel (lead is a superior gamma shield, but has a low effectiveness in shielding invessel structures from fast neutrons.) Flowing lead in the downcomer between the shroud and the reactor vessel is also modelled in the neutronics analysis. 

Two hundred and ninety drawn glass beads and fragments of varying colour, structure (flashed, cored and uncored), sizes, and shapes from site Asd/KglO were analyzed non-destructively using instrumental neutron activation analysis at the SLOWPOKE Reactor Facility of the University of Toronto. These beads had to be neutron-irradiated as little as possible in order to minimize the build-up of radioactivity from Sb (half-life 2.75 days) and from Sb (half-life 60.9 days) in the Sb-rich beads, so that they could be returned to their owners within a reasonable amount of time. 

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