On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO₂ fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.     

Advanced methods in teaching reactor physics

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

On criticality calculations in multislab geometry

The spherical-harmonics method is used to solve a class of multigroup criticality (k-eigenvalue) problems in multislab geometry. The model includes scattering anisotropy of arbitrary order and allows reflective or vacuum outer boundaries. Numerical results of benchmark quality are reported for three sample problems that have been defined and used by other authors to study various transport methods for criticality calculations. A comparison with our results indicates that some of the earlier results are in error.     

A new physics design of control safety rods for prototype fast breeder reactor

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B₄C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B₄C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B₄C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B₄C sections in top and bottom. There is significant savings in the initial inventory of enriched B₄C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles.     

Two-phase flow phenomena in full-scale reactor geometry

For the investigation of two-phase flow phenomena in full scale reactor geometry, a series of experiments were carried out at the Upper Plenum Test Facility UPTF, which represents the primary system of a 1300 MWe Pressurized Water Reactor with upper plenum, downcomer and primary main coolant pipes in 1:1 reactor scale.UPTF was the German contribution to the international 2D/3D project established by the Japan Atomic Energy Research Institute (JAERI), the Nuclear Regulatory Commission (USNRC) of the United States of America, and the Federal Ministry for Research and Technology (BMFT) of the Federal Republic of Germany.Large scale findings of the UPTF tests, related to two-phase flow phenomena in the downcomer, in the upper plenum, at the upper core tie plate, and in the main coolant pipes, will be discussed. The application of the UPTF test results for the validation of analytical models will be demonstrated.     

Experience in using the finite element method for reactor calculations

The finite element method can be used to solve the stationary and the transient neutron diffusion equation. Formulations and discretizations of both equations are given. Experience derived from applying the finite element method to practical reactor physics problems is summarized.     

Propagation noise calculations in VVER-type reactor core

Neutron noise induced by propagating disturbances in VVER-type reactor core is addressed in this paper. The spatial discretization of the governing equations is based on the box-scheme finite difference method for triangular-z geometry. Using the derived equations, a 3-D 2-group neutron noise simulator (called TRIDYN-3) is developed for hexagonal-structured reactor core, by which the discrete form of both the forward and adjoint reactor dynamic transfer functions (in the frequency domain) can be calculated. In addition, both types of noise sources, namely point-like and traveling perturbations, can be modeled by TRIDYN-3. The results are then benchmarked in different cases. Considering the noise source as propagating perturbations of the macroscopic absorption cross sections, the induced neutron noise is calculated throughout the reactor core. For the first time, adjoint approach is applied and examined for modeling moving noise sources. Moreover, the space- and frequency-dependence of the propagation noise are investigated in this paper.     

Recent activities in the field of reactor physics

ABSTRACT

Reactor Physics that treat the essentials of how fission nuclear reactors work fundamentally has played important roles in safe operations and design studies of various types of nuclear reactors. From the latest activities in the field of reactor physics, this report summarizes some outstanding researches and developments published in scientific journals, including the Journal of Nuclear Science and Technology and others.     

快堆三维节块法程序燃耗模块的开发

快堆三维节块法程序(HND)由于其在计算精度和计算时间上的优越性,在中国实验快堆CEFR的初步设计和施工设计中都起了很大的作用。但是,三维节块法本身不具备燃耗计算的功能。本文的主要内容是对三维节块法添加燃耗计算的模块,并采用IAEA基准例题和中国实验快堆的对比计算进行检验。记算带有燃耗模块的HND为HNDB,结果表明,HNDB在燃耗计算中有较好的计算结果。     

A mathematical model and operational program complex for neutron physics processes in the BN-350 reactor

Translated from Atomnaya Énergiya, Vol. 69, No. 6, pp. 368–370, December, 1990