Volume 80, Issue 6

TRACE is used to calculate the thermal-hydraulic sequence in nuclear power plants for accident analysis. In some postulated accidents, countercurrent flow limitation (CCFL) phenomena can occur. This phenomenon is calculated by means of empirical relationships in TRACE. Usually, not all empirical relationships used in TRACE are accessible to the user who develops the computational model for accident analysis, but in the case of CCFL, the user must specify the parameters of the empirical relationship. Additional preliminary work is needed for fitting the parameters so that experimental data matches the computational calculations and the accident analysis can be performed. Furthermore, uncertainties in measurement errors from experimental data lead to user-based variations in the parameters of the empirical relationship for CCFL. Therefore, an alternative procedure in which accident analysis is performed with consideration of the uncertainties in the empirical relationships is presented. The uncertainty was quantified by means of a stochastic linear regression model for the Wallis correlation and the Bankoff correlation. An uncertainty analysis was subsequently performed to demonstrate the influence of uncertainty in certain cases of CCFL.

Lungmen nuclear power plant project started long time ago, it is not yet commercially operated but Taiwan Power Company has already prepared for its startup tests. 3RIP trip startup test is one of them. Three of the 10 RIPs will be manually tripped in the test. Response of the plant for this transient will be watched and recorded to check if the test criteria are satisfied. This paper is a result of code simulation of 3RIP trip startup test of Lungmen ABWR nuclear power plant. Thermal hydraulic code TRACE coupled with neutronics code PARCS were used to build the simulation model of Lungmen nuclear power plant. Startup tests under different plant power and flow conditions are considered in this research. A sensitivity study on the impact of different pump moment of inertia has been performed. Simulation results with TRACE and PARCS shows that the acceptance criteria for this startup test can be satisfied and the impact of different pump inertia is little.

In this paper, we developed a new approach of analytic function expansion nodal (AFEN) method to solve the multi-group and multi-dimensional neutron diffusion equation in reactor cores with hexagonal fuel assembly. This method represents a multidimensional intra nodal flux distribution in terms of analytic basis functions at any points in the node. New types of boundary conditions have been considered that constrain the intranodal flux distributions in the hexagonal-z node, which include twelve radial surface-averaged partial currents and two axial surface-averaged partial currents. We utilized the coarse group rebalancing (CGR) method to increase the speed of code calculations. The computer code takes a few-groups cross sections produced by a lattice code and calculates the effective multiplication factor ( k eff ), flux in multi-group energy, reactivity, and the relative power density at each fuel assembly. Finally, the solution accuracy is tested for two well-known benchmark problems. The numerical results demonstrate that the new AFEN method is an accurate method for calculating k eff and power density distribution in hexagonal-z geometries.

As a physical method, laser cleaning technology in equipment maintenance will be a good prospect. The experimental apparatus for laser cleaning of heat tubes in the steam generator was designed according to the results of theoretical analysis. There are two conclusions; one is that laser cleaning technology is attached importance to traditional methods. Which has advantages in saving on much manpower and material resource and it is a good cleaning method for heat tubes. The other is that the acoustic emission signal includes lots of information on the laser cleaning process, which can be used as real-time monitoring in laser cleaning processes. When the laser acts for 350 s, 100 % contaminants of heat tubes is cleaned off, and the sensor only receives weak AE signal at that time.

The High-Performance Light Water Reactor (HPLWR) is the European version of the advance nuclear energy power plant at Supercritical-pressure. A light water reactor at supercritical pressure, being currently under design, is the new generation of nuclear reactors. The aim of this study is to predict the HPLWR neutronic behavior of the axial enrichment profile with an average enrichment of 5 w/o U-235. Neutronic calculations are performed using WIMS and CITATION codes. Changes in neutronic parameter, such as Power Peaking Factor (PPF) are discussed in this paper.

In this paper, the analysis of a Small Break Loss of Coolant Accident (SBLOCA) in the VVER-1000/V446 nuclear power plant is presented. For a conservative analysis of the accident, the loss of power to the NPP and failure of one accumulator, and also of two emergency core cooling systems (ECCS) in loops 2 and 3 of the primary and secondary circuits are considered when SBLOCA has occurred. The RELAP5/MOD3.2 computer code has been used in performing the analyses. Two cases of accident scenarios as 25 mm and 100 mm breaks are analyzed. The results are in good agreement with those reported in the plant's FSAR. The results of liquid velocity show that in both cases, the flow of hot legs after the break is reversed, which provides the potential for reflux condensation phenomena. Furthermore, in the 25 mm break, the flow rate in the broken and intact side downcomer remains in the downward motion while in the 100 mm break, the broken and intact side flow rate changes to the reversed state alternatively.

The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO 2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (“Onion-Skin” effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

51 Cr (T 1/2 = 27.7 d), which decays via electron capture (100 %) with 320 keV gamma emission (9.8 %), is a radionuclide with still a large application in biological studies. In this work, ALICE/ASH and TALYS nuclear model codes along with some adjustments are used to calculate the excitation functions for proton, deuteron, α-particle and neutron induced on various targets leading to the production of 51 Cr radioisotope. The production yields of 51 Cr from various reactions are determined using the excitation function calculations and stopping power data. The results are compared with corresponding experimental data and discussed from point of view of feasibility.

Radio isotopic tracer technique as one of the versatile nondestructive technique is employed to evaluate the performance of nuclear grade anion exchange resins Tulsion A-23 and Dowex SBR LC. The evaluation was made on the basis of ion-isotopic exchange reaction kinetics by using 131 I and 82 Br radioactive tracer isotopes. It was observed that for both the resins, the values of specific reaction rate (min −1 ), amount of ion exchanged (mmol) and initial rate of ion exchange (mmol/min) were calculated to be lower for bromide ion-isotopic exchange reaction than that for iodide ion-isotopic exchange reaction. It was observed that for iodide ion-isotopic exchange reaction under identical experimental conditions of 30.0 °C, 1.000 g of ion exchange resins and 0.001 mol/L labeled iodide ion solution, the values of specific reaction rate (min −1 ), amount of iodide ion exchanged (mmol), initial rate of iodide ion exchange (mmol/min) and log K d were calculated as 0.377, 0.212, 0.080 and 15.5 respectively for Dowex SBR LC resin, which was higher than 0.215, 0.144, 0.031 and 14.1 respectively as that obtained for Tulsion A23 resins. Also at a constant temperature of 30.0 °C, as the concentration of labeled iodide ion solution increases from 0.001 mol/L to 0.004 mol/L, the percentage of iodide ions exchanged increases from 84.75 % to 90.20 % for Dowex SBR LC resins which was higher than increases from 57.66 % to 62.38 % obtained for Tulsion A23 resins. The identical trend was observed for the two resins during bromide ion-isotopic exchange reaction. The overall results indicate superior performance of Dowex SBR LC over Tulsion A23 resins under identical experimental conditions.

One speed, time independent neutron transport equation in slab geometry can be solved with the triplet anisotropic scattering for the criticality problem with the reflected boundary condition. The value of the critical slab thickness is investigated numerically by using the Modified F N (H N ) method. Case's eigenfunctions, normalization relation and orthogonality relations must be derived for this scattering in order to use the Modified F N method. Some selected values, which can be calculated from the criticality equation, can be tabulated. Also the term which comes from the Poincaré-Bertrand formula may be added to the orthogonality formula of the continuum eigenfunction.