Ex-core Detector Response Research Articles

Secondary source reconfiguration for a Westinghouse four-loop plant

An approach is described for estimating the excore source range detector count rate for several secondary neutron source configurations. A case in which a Westinghouse four-loop plant considered relocating the secondary source fuel assemblies and switching from secondary source fuel assemblies with six secondary source rods to assemblies with two or four secondary source rods is presented. The excore detector count rate is first estimated computationally, by modeling the secondary source fuel assemblies and excore detectors in MCNP and predicting the detector count rate based on detector sensitivity in cps/nv listed in the detector manual. Since the excore detector response can vary greatly from cycle to cycle even for equilibrium loading patterns due to variations in detector response, benchmarking to actual excore source range detector count rates is presented. A correction factor for each individual source range detector is determined from the benchmark, then applied in the prediction of excore source range detector count rate for the proposed two or four secondary source rod configurations. The approach presented in this paper could also be used to perform an engineering evaluation on the feasibility of elimination of secondary sources.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

A HYBRID MONTE-CARLO-DETERMINISTIC METHOD FOR AP1000 EX-CORE DETECTOR RESPONSE SIMULATION

The ex-core detector-response calculation is a typical deep-penetration problem, which is challenging for the Monte Carlo method. The response of the ex-core detector is an important parameter for the safe operation of the nuclear power plants. Meanwhile, evaluation of the ex-core detector response during each step of fuel-loading is used to guide the fuel-loading sequence. The response can also be used to reconstruct core-power distribution for online monitoring of long-term power. The detector used for the ex-core response is the source-range detector which is sensitive to thermal neutrons. For a Monte Carlo shielding calculation of the above detector response, the thermal flux under 0.625eV is needed, which is too small to be tallied by traditional Monte Carlo simulations. In practice, the tally results are close to zero in the detector region under direct Monte Carlo calculation. Even if the number of particles is increased to a significant amount, the statistical variance is still very large. The high variance along with a significant calculation time leads to a small Figure Of Merit (FOM). In order to solve this problem and to improve the tally efficiency of the ex-core detector response, a hybrid Monte-Carlo-deterministic method is employed in this study, and an in-house hybrid Monte-Carlo-deterministic particle transport code, NECP-MCX, is developed in this paper. The method takes the space-energy-dependent adjoint fluxes to generate importance parameters for the mesh-based weight window in the Monte Carlo calculation. Simultaneously, the mesh-based source biasing is performed with the consistent importance parameters to make the starting weight of neutrons matching with the survival weight of the weight windows. As the mesh used in the hybrid Monte-Carlo-deterministic method is superimposed, the mesh of the weight window will not be affected by the complex geometry model. The adjoint flux is obtained by the efficient SN method with the multi-group cross-section data. The whole toolset is convenient to use with single set of the modelling data for both Monte Carlo and deterministic simulations. Compared with the direct Monte Carlo simulation, the hybrid Monte-Carlo-deterministic method has a higher efficiency for a typical deep-penetration problem such as the AP1000 ex-core detector-response simulation.

Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores

In this paper, the ex-core source-range detector response was simulated for AP1000 cores in its fuel-loading process by using a Monte Carlo code. Firstly, the current sensitivity coefficients of the ex-core source-range detectors have been calibrated by utilizing the experiments with single fuel assembly and the primary source loaded in the core. Secondly, both relative and absolute calibration methods have been applied and analyzed for the sensitivity coefficients calibration. Thirdly, the simulation values for the ex-core source-range response with fully loaded fuel have been validated by using the actual measurement of the first AP1000 cores in the world. Fourthly, the contribution of the 238U spontaneous fission to the ex-core source-range detector response has been studied. Through the numerical simulation results, the following observations can be obtained: 1) The relative calibration method is better than the absolute calibration method, especially when the ex-core source-range response is small. 2) Compared with the measurements, the relative deviation between the simulated and the measured values is within 14% for the two AP1000 cores at Sanmen Nuclear Power Plant. 3) The intensity of 238U spontaneous fission source accounts for 15% of the total source strength in the AP1000 core Unit 1, but the maximum change of response is only 0.02 cps, which is still insignificant for the ex-core detector response simulation in PWR.

Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Study of the PWR ex-core detector response simulation based on the 3D SN method

Simulation of the ex-core detector response function (DRF) is of great importance for reactor power monitoring, control and protection systems. However, the 3D DRF calculation requires vast majorities of computational resources due to its characteristics such as deep penetration, strong anisotropic scattering, large computational geometry, etc. In this paper, the step characteristics (SC) spatial differencing scheme and GMRES algorithm have been implemented in the 3D SN code to ensure the simulation efficient. The adjoint first-collision (FC) source method is developed to mitigate the ray effect in the DRF simulation. The Kobayashi benchmarks are calculated to verify the correctness of the FC module. Numerical results based on the Kori-1 reactor ex-core detector indicate that the SN-FC method has the comparable accuracy with multi-group Monte-Carlo method but achieves much higher efficiency.

Three-dimensional distribution calculation for ex-core detector response functions

Three-dimensional distribution calculation for ex-core detector response functions

Comparisons of SN and Monte-Carlo methods in PWR ex-core detector response simulation

The ex-core detector response calculation is an important part in reactor design. However, the response function cannot be measured by experiments quantitatively. Ex-core detector response simulation is therefore required. For decades, the SN code has been used as the dedicated tool. Nowadays, more and more engineers are expressing an interest in using the Monte-Carlo method instead of the SN method in simulations, as it is expected that the Monte-Carlo method will give higher accuracy. In this paper, the modeling and simulation of ex-core detector responses is briefly reviewed based on the Korean Kori Unit 1 reactor. Then, the differences between the SN simulation and Monte-Carlo simulation are compared. The sensitivity of computational conditions is also discussed. It is shown that the problem dependence of cross sections and meshing dependence of spatial discretization in the ex-core detector response calculations are not as strong as expected. However, the ray effect is the main shortcoming for the SN calculation. Based on the analysis, two benefits are shown by using MCNP for the direct 3D calculation. Firstly, the impact of ray effect is eliminated without using the SN angular discretization. Secondly, the direct 3D calculation is easier to perform based on the powerful ability of 3D modeling and parallel computing of the Monte-Carlo code. The new DRF values are adopted in the dynamic control rod reactivity measurement of Kori Unit 1 reactor. The results show that the new DRF values improve the error of measured control rod worth by a percentage of 3.

A preliminary simulation study of dynamic rod worth for the SMART (System-integrated Modular Advanced ReacTor) reactor

This paper describes a preliminary simulation of the dynamic rod worth for the SMART (System-integrated Modular Advanced ReacTor) reactor core which has been developed by KAERI (Korea Atomic Energy Research Institute). This work was done by comparing the dynamic rod worth calculated using a simple inverse kinetic model with simulated pseudo ex-core detector signals and the static rod worth calculated with 3D static core diffusion nodal calculations. The kinetic parameters for the simple inverse kinetic model are estimated by using 3D static core diffusion nodal calculations as functions of the rod positions while the pseudo ex-core detector signals as the control rod positions change are simulated by using 3D transient core diffusion calculations with the MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors) code and the ex-core detector response function calculated with the DORT code. Then, the pseudo ex-core detector signals are converted into the core neutron flux using the pre-estimated ex-core detector response conversion factor (DRCF). The simulation results show that the dynamic rod worth with the simulated pseudo ex-core detector signals have good agreement with the static rod worth for the SMART reactor and that the deviations of the static and dynamic rod worth from uncertainties of control rod group constants show good agreements with each other. So, it is considered that this work can provide baseline data for the actual simulation using real ex-core detector signals for the SMART reactor in the future.

Locally optimal solution of robust ex-core detector response using constrained simulated annealing

This note describes the determination of the parameter vector used to estimate the axially 3-level average core power with ex-core detectors. The theoretical result of the previous work is performed by the well-known constrained quadratic programming method with active set method. Sometimes the previous method gives the solutions on the constraint boundaries not far from the inequality constraints although the solutions are feasible and acceptable. In this note, a constrained simulated annealing method is proposed to find the locally optimal but robust solutions. The method is applied to plant measurement data and its performance is demonstrated by core follow calculation for a core depletion cycle. The test results show the improvement in finding solutions not stuck on constraint boundary and its robustness by reconstruction accuracy for saddle type power distributions using the noisy measurement data.

Ex-Core Detector Response Evaluation of the SMART Reactor by Using the DORT Code

The development of a 660 MWt integral reactor SMART (System integrated Modular Advanced ReacTor), called SMART-660, has been initiated in Korea. A preliminary evaluation of the detector response functions was performed for the ex-core detector of the SMART-660. First, a candidate position to install the ex-core detector was selected by considering the reactor configuration, the detector performance, and the thermal neutron flux distribution in the reactor assembly. Second, the detector response functions such as the shape annealing function (SAF) and assembly weighting factor (AWF) for the candidate ex-core detector were calculated by using an adjoint transport calculation with the DORT code and the BUGLE-96 cross-section library. The GEOSHIELD program was used to generate the DORT input file and to process the DORT output file with a graphic visualization. This study has shown that the response functions of the ex-core detector are only sensitive to some parts of the reactor and additional efforts are required to reduce the locality of the ex-core detector response.

Ex-Core Detector Response Caused by Control Rod Misalignment Observed during Operation of the Reactor on the Nuclear ShipMutsu

Unexpected deviations of ex-core neutron detector signals were observed during a voyage of the Japanese nuclear ship, Mutsu. From detailed three-dimensional analyses, this phenomenon was determined to be caused by an asymmetrical neutron source distribution in the core due to a small misalignment between the two control rods of a control rod group. A systematic ex-core detector response experiment was performed during the Mutsu's third experimental voyage to gain some understanding of the relationship between the control rod pattern and the detector response characteristics. Results obtained from analyses of the experiment indicate that the Crump-Lee technique, using calculated three-dimensional source distributions for various control rod patterns, provides good agreement between the calculated and measured detector responses. Xenon transient analyses were carried out to generate accurate three-dimensional source distributions for predicting the time-dependent detector response characteristics. Two types of ex-core detector responses are caused by changes in the control rod pattern in the Mutsu reactor: the detector response ratio tends to decrease with the withdrawal of a group of control rods as a pair, and a difference in the positions of the control rods in a group causes signal deviations among the four ex-core detectors. Control rod misalignment does not greatlymore » affect the mean value of the four detector signals, and the deviation can be minimized if the two rods within a group are set at the same elevation at the time of detector calibration.« less

Analysis of Ex-Core Neutron Detector Response during a Loss-of-Coolant Accident

In this paper the experimental response of ex-core neutron detectors during both actual and simulated loss-of-coolant accidents (LOCAs) at a pressurized water reactor are analyzed to determine their cause. Various analytical techniques are used to reproduce the ex-core detector response during large-break LOCAs. These techniques include both discrete ordinates transport and point kernel calculations. The experiments analyzed include large-break LOCA experiments at the Loss of Fluid Test Facility and from the Three Mile Island accident. The results show that an adiabatic method is sufficiently accurate to reproduce the detector response. This response can be explained in terms of the combined effects of changes in shielding and multiplication that occur in a core during a LOCA.

Analytical method for recalibration of incore and excore nuclear instrumentation system of PWRs.

In pressurized water reactors (PWRs) of nuclear power plants, it is periodically required to recalibrate the correlation between the incore and excore nuclear instruments, where the recalibration guarantees the accurate monitoring of incore axial power distributions by excore instrumentation. In order to obtain the correlation between incore power distribution and excore detector response, however, it is necessary to perturb the axial power distribution of the reactor by control rods or to introduce a xenon spatial oscillation. In this paper a simplified analytical method that eliminates such perturbations is proposed to evaluate the correlation between incore power distribution and excore detector currents which is utilized to recalibrate the nuclear instrumentation system. This method provides as good an accuracy as that obtained by conventional methods.

Analysis of Ex-Core Detector Response Measured During Nuclear ShipMutsuLand-Loaded Core Critical Experiment

There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment.

Reevaluation of Spatial Weighting Factors for Ex-Core Neutron Detectors

The ex-core detector response in pressurized water reactors (PWRs) depends on not only power level but also core power distribution. Therefore, it is important to precisely calculate the assembly-wise spatial weighting factors for the ex-core detectors. Usually these factors are calculated with the one-dimensional transport code and point kernel calculational method, in which the neutron scattering effect outside of reactor vessel is neglected. But when the scattering effect is estimated to be rather big, we calculate the assembly-wise spatial weighting factors using the two-dimensional transport code, which includes the scattering effect. Consequently, we found that the weighting factors of peripheral assemblies that are remote from the detector but close to reactor vessel are rather big in comparison with the previous results. When we calculate the detector response during one control rod insertion test of three-loop PWR core using these weighting factors, the agreement between calculation and measurement is very good. A simple point kernel calculational method developed instead of the two-dimensional transport calculation that consumes much computer time.

Calculation of Spatial Weighting Functions for Ex-Core Neutron Detectors

A mathematical model for ex-core detector response in pressurized water reactor (PWR) configurations is presented, based on the use of a spatial weighting function that is independent of core power...