Thermal Reactor Physics

R. Chawla, An Assessment of Methods and Data for Predicting Integral Properties for Thermal Reactor Physics Experiments, AEEW-R797 (1972). 

Safely implementing a thermochemical nuclear hydrogen generation scheme requires a robust understanding of the interaction between the nuclear plant and the chemical plant. In turn, this requires robust models of the chemical plant, reactor thermal-hydraulics and reactor physics. Efforts have been conducted in both the transient modelling of the sulphur-iodine (S-I) and hybrid sulphur (HyS) thermochemical cycles, as well as coupling to models of the pebble bed modular reactor (PBMR-268) (Brown, 2009). 

In this paper we will discuss the application of a general batch reactor model that considers the reaction kinetics, heats of reaction, heat transfer properties of the reactor, physical properties of the reactants and the products, to predict 1) The concentration profile of the products, thus enabling process optimization 2) Temperature profile during the reaction, which provides a way to avoid conditions that lead to a thermal runaway 3) Temperature profile of the jacket fluid while maintaining a preset reactor temperature 4) Total pressure in the reactor, gas flow rates and partial pressure of different components. The model would also allow continuous addition of materials of different composition at different rates of addition. 

Thus the constant of proportionality has decreased by about 30% over the life of the core. More accurate relationships between thermal power density and neutron density at different stages of the reactor run may be available from the design calculations or from plant-specific reactor physics data. 

The choice of cladding material for fast reactors is less dependent upon the neutron absorption cross section than for thermal reactors. The essential requirements for these materials are high melting point, retention of satisfactory physical and mechanical properties, a low swelling rate when irradiated by large fluences of fast neutrons, and good corrosion resistance, especially to molten sodium. At present, stainless steel is the preferred fuel cladding material for sodium-cooled fast breeder reactors (LMFBRs). For such reactors, the capture cross section is not as important as for thermal neutron reactors. 

On the basis of theoretical calculations by the Physics Division, the Technical Division of Clinton Laboratories submitted a design proposal in May, 1946 for a high-flux thermal reactor of heterogeneous core, light water—cooled and heavy watei moderated and—reflected. The use of light water as a coolant was believed preferable to the use of heavy water because light water increased the operation flexibility of the machine. A brief description of this proposal is given. 

R.D. Lonsdale, "An algorithm for solving thermal-hydraulic equations in complex geometries The ASTEC code", in Proceedings of International Topical Meeting on Advances in Reactor Physics, Mathematics and Computation, pp. 1653-1664, Paris, France, 27-30 April, 1987. 

Even if the full details of the dynamics of moderator atom motion were known, it would not be possible to incorporate all this information into neutron thermalization calculations for reactors. Therefore, it is necessary to work with somewhat simplified models. Especially in the absence of a complete knowledge of the appropriate differential cross sections, it is necessary to compare calculations using these models with clean experimental information on the neutron distribution in position and energy for configurations of interest in reactor physics. 

Fermi must have doubted the potential of thermal reactors for future energy production as he proposed a fast reactor (FR) for this purpose (Fermi 1944) - a reactor with a large excess of neutrons over the amount required for maintaining a chain reaction. Neutron excess (NE) is the key physical resource for dealing with the problems of fuel and, as is apparent now, of safety, given an adequate choice of technological tools ... 

The measurements mentioned have been used for two purposes to develop semiempirical calculatlonal schemes, recipes , for the design of real reactor and to provide experimental information with which theory can be compared, ip the hope of arriving, eventually, at a real understanding bf the physics of thermal reactors. A serious problem that arises is that of the possible difference between results of measurements of a particular quantity when made in a critical assembhf and in a subcritical assembly. For example, recent work at the Savannah River liaboratory has indicated significant differences between values of the material buckling obtained from exponential and critical assemblies moderated by heavy water. A similar effect does not seem to have been observed in assemblies moderated by graphite or ordinary water. 

F. FEINER and L. J. ESCH, "Survey of Capture and Fission Integrals of Fissile hlaterials," Reactor Physics in the Resonance and Thermal Regions, Am. Nucl. Soc. Nat l. Topical Mtg., San Diego (Febi ry (1966). 

F. R. NAKACHE and S. KELLMAN, Operating Experience with UNC-THERMOPILE, an Advanced Monte Carlo Program for the Evaluation of Thermal Assemblies, Proc. Conf. Reactor Physics in the Resonance and Thermal Regions (Feb. 1966). To be published. 

F. FEINER, Proc. Am. Nucl. Soc., Topical Meeting Reactor Physics in Thermal and Resonance Regions (February 1966). 

J. J. SCHMIDT, "Resonance Properties of the Main Fertile and Fissionable Nuclei," Reactor Physics in the Resomnee and Thermal Regions, Vol. II, pp. 223-260, Proc. Natl. Topical Meeting Am. Nucl. Soc., San Diego (February 7-9, 1966). 

J, R. ASKEW, Some Problems in the Calculation of Resonance Capture in Lattices, ANS National Topical Meeting on Reactor Physics in the Resonance and Thermal Regions, San Diego (February 1966). 

L. C. SCHMID, B. R. LEONARD, Jr., R. L. LIIKALA, and R. I. SMITH, "Reactor Physics Data for the Utilization of Plutonium in Thermal Power Reactors," BNWL-801, Pacific Northwest Laboratory (May 1968). 

Similar to other thermal reactor designs, there are three basic functions that are necessary to mitigate the consequences of fission product releases during a postulated accident. The functions, referred to as the 3 Cs, are control, cool, and contain. Control refers to safe reactor shutdown. Cool involves the removal of heat—from the fuel produced by the fission process (at power) or by the decay heat after reactor shutdown—and rejection of the heat to a heat sink. Contain is simply the physical means to prevent the release of radioactive material to the atmosphere by provision of containment systems. 

Loss of regulation Fuel sheath HT system boundary Reactor physics System thermal hydraulics... 

Loss of HT flow Fuel sheath HT system boundary System thermal hydraulics Reactor physics... 

Laslau, P. and D. Serghiuta. 1991. Reactor Physics and Thermal Hydraulic Calculations, Internal Report INR 3360, Pitesti, Romanian. 

Die x sics asspclated vlth a mlt lattice cell is a convenient and useful -means of describing the reactor physics of a large thermal reactor In this section the reactor vill be assumed to be infinite in extent and each lattice cell therefore is one of an infinite set of identical cells The modifications to such a model are discussed in the next section on core physics ... 

In the system under consideration (a noncirculating-fuel bare thermal reactor), it is physically reasonable to suppose that the spatial variation of the concentration of delayed-neutron precursors is proportional to that of the neutron flux and that this mode persists even though the magni-... 

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