Nuclear Power Institute Of China Research Articles

Assemblies design and modeling analysis of a new fine mesh neutronics/thermal-hydraulics coupling benchmark for plate-type PWR core

In order to support the verification of neutronics/thermal-hydraulics coupling calculation method or simulation codes at the fine mesh for plate-type pressurized water reactor (PWR) with high parameters (which the ratio of power to mass flow rate is greater than 235 kW/kg, the core outlet enthalpy exceeds 1500 kJ/kg), a set of coupling calculation of plate-type PWR based on high parameters (COPHP) is design by Nuclear Power Institute of China (NPIC). Multiple industry research teams, including Tsinghua University, Xi'an Jiaotong University, and Huazhong University of Science and Technology, participated in the production of COPHP benchmark. This article provides a detailed explanation of the completed assembly design and modeling calculations, and provides the keff, pin-by-pin wise fine mesh relative power distribution, and deviation results calculated for 10 conditions of 6 assemblies using RMC, OpenMC, and KYLIN V2 software. The results show that: 1) when using the same cross-sectional library, the OpenMC calculation results are in good agreement with the RMC results. Taking ENDF/B-VII.1 as an example, the maximum deviation of keff in the entire burnup process of all assemblies is −157pcm, the maximum deviation of relative power is −1.13 %, and the maximum power weight error(PWE) is 0.226 %. 2) Compared between KYLIN V2 and the RMC by using ENDF/B-VII.1, the maximum deviation of keff is −468pcm, the maximum deviation of relative power is −1.49 %, and the maximum PWE is 0.316 % when calculating the standard assemblies with all control rod out(ARO) condition and two burnable poison assemblies. The maximum deviation of keff is −795pcm, the maximum deviation of relative power is 1.81 %, and the maximum PWE is 0.369 % when calculating the standard assemblies with all control rod inserted(ARI) conditions.

Validation of whole core transport code SHARK with HPR1000 experiment

SHARK is a new whole core transport code developed by the Nuclear Power Institute of China (NPIC). The constructive solid geometry (CSG) is used in SHARK to model the reactors. The subgroup method based on an equivalence cross-section interpolation table is applied to treat resonance self-shielding effects. The 2D/1D transport methods can be chosen in SHARK. The HPR1000 which is a 3rd commercial reactor has independent intellectual property owned by China. The validation of SHARK is processed via the HPR1000 in this work. The numerical results demonstrate that SHARK can calculate whole core directly and accurately.

Subgroup method for the high fidelity neutronics code SHARK and preliminary benchmarking

SHARK (Simulation-based High-fidelity Advanced Reactor physics Kit) is a numerical reactor software newly developed by Nuclear Power Institute of China (NPIC) based on the full core deterministic heterogeneous transport theoritical framework. SHARK uses subgroup method to characterize the complex heterogeneous resonance self-shielding effects based on a 45-group library. Both 2D/1D Method of Characteristics (MOC) and quasi-3D MOC are developed as the one-step heterogeneous transport solvers. This program is still under development, and recently the ability of steady microscopic-cross-section level problem calculations is established and verified. This paper emphasizes the subgroup methodology and related improvements in SHARK program, which has decisive influences on the final accuracy. The subgroup method applied in this paper can significantly reduce the cost of transport computations from the classical one, which will benefit the heterogeneous core calculations. Based on VERA core physics benchmark problems, the program is preliminarily verified. The results basically show good agreements with the reference solutions, which reveals the simulating accuracy level of the SHARK program for commercial Pressurized Water Reactor (PWR) assemblies.

FSAR: Full neutron spectrum code for advanced reactor simulation

The full neutron spectrum code for advanced reactor simulation named FSAR has recently been developed at Nuclear Power Institute of China in order to meet the requirements of advanced reactor with large neutron energy range. Based on the two-step calculation scheme, FSAR consists of two-dimensional lattice calculation code and core calculation code. In two-dimensional lattice calculation, the subgroup method with ultrafine energy groups was implemented in the two-dimensional resonance self-shielding by incorporating the MOC to get the accurate self-shielded cross-section. For better consideration of the strong space coupling in different geometric sizes due to different mean free path of neutrons, the super-homogenization method and the leakage model were applied. In the core calculation, the discrete-ordinate method and micro burnup calculation method were used to simulate the core neutron transport and depletion. Preliminary calculation results showed that for the problems with wide spectrum, the self-shielded cross-sections have a good agreement with the Monte Carlo solution. The results shown in this paper indicate that FSAR has good performances of the cross-section generation in full neutron spectrum problem simulation.

Monte Carlo dosimetry of the 60Co sources of a new GZP3 HDR afterloading system

BackgroundThe purpose of this work was to obtain the dosimetric parameters of the new GZP3 60Co high-dose-rate afterloading system launched by the Nuclear Power Institute of China, which is comprised of two different 60Co sources.MethodsThe Monte Carlo software Geant4 and EGSnrc were employed to derive accurate calculations of the dosimetric parameters of the new GZP3 60Co brachytherapy source in the range of 0–10 cm, following the formalism proposed by American Association of Physicists in Medicine reports TG43 and TG43U1. Results of the two Monte Carlo codes were compared to verify the accuracy of the data. The source was located in the center of a 30-cm-radius theoretical sphere water phantom.ResultsFor channels 1 and 2 of the new GZP3 60Co afterloading system, the results of the dose-rate constant (Λ) were 1.115 cGy h−1 U−1 and 1.112 cGy h−1 U−1, and for channel 3 they were 1.116 cGy h−1 U−1 and 1.113 cGy h−1 U−1 according to the Geant4 and EGSnrc, respectively. The radial dose function in the range of 0.25–10.0 cm in a longitudinal direction was calculated, and the fitting formulas for the function were obtained. The polynomial function for the radial dose function and the anisotropy function (1D and 2D) with a uptheta of 0°–175° and an r of 0.5–10.0 cm were obtained. The curves of the radial function and the anisotropy function fitted well compared with the two Monte Carlo software.ConclusionThese dosimetric data sets can be used as input data for TPS calculations and quality control for the new GZP3 60Co afterloading system.

Study on the calculational framework development of the advanced numerical reactor neutronics code SHARK

The SHARK program (Simulation-based High-fidelity Advanced Reactor physics Kit) is a high-fidelity heterogeneous neutronics code for the numerical reactor system being developed at the Nuclear Power Institute of China (NPIC). The program uses a Constructive Solid Geometry (CSG) framework to model various complex geometries. To enhance the flexibility and robustness during continuous development process, SHARK program attempts to support a rich set of methods, tools and library options within a unified general framework as a “toolkit”. For the multi-core clustered supercomputer architectures that are commonly used today, the SHARK program adopts a hybrid parallel strategy of MPI and OpenMP to achieve complementary advantages between them. In addition, the framework of SHARK program is designed with a true object-oriented manner. Through reasonable abstraction, inheritance and encapsulation, the maintainability and extensibility of the code are improved, and long-term team development is facilitated. Up to now, key modules for cross-section generation, heterogeneous transport calculation and microscopic depletion have been developed under the general frameworks. The main features of SHARK’s “resonance-transport-depletion” coupling system are elaborated in this paper, and some verification and validation (V&V) results in the current phase are presented and discussed.

Parameter identification and state estimation for nuclear reactor operation digital twin

Reactor Operation Digital Twin (RODT) is now receiving increasing attention and investment in nuclear engineering domain. A prototype of a RODT was first brought out by Gong et al. at Nuclear Power Institute of China. The RODT contains a forward solver for online real-time simulation and an inverse problem solver for parameter identification and state estimation. To further improve the efficiency and accuracy of RODT and promote the practical deployment of RODT (i) we first propose an advanced differential evolution algorithm to upgrade the inverse solver; (ii) then we bring out a systematical uncertainty quantification of RODT considering assimilation noisy observations. The accuracy and validity of the proposed RODT are tested along one cycle operating stage of HPR1000, at various operating temperatures, control rod steps, general power level and the inlet temperature of a practical nuclear reactor core. Numerous numerical results confirm its potential for practical engineering applications for on-line parameter identification and state estimation.

Optimization of preventive maintenance of nuclear safety-class DCS based on reliability modeling

Nuclear safety-class DCS is used for nuclear reactor protection function, which is one of the key facilities to ensure nuclear power plant safety, the maintenance for DCS to keep system in a high reliability is significant. In this paper, Nuclear safety-class DCS system developed by the Nuclear Power Institute of China is investigated, the model of reliability estimation considering nuclear power plant emergency trip control process is carried out using Markov transfer process. According to the System-Subgroup-Module hierarchical iteration calculation, the evolution curve of failure probability is established, and the preventive maintenance optimization strategy is constructed combining reliability numerical calculation and periodic overhaul interval of nuclear power plant, which could provide a quantitative basis for the maintenance decision of DCS system.

Validation of the advanced neutron transport lattice code KYLIN-V2.0 with critical and burnup experiments

KYLIN-V2.0 is an advanced lattice code developed in Nuclear Power Institute of China. The subgroup method, MOC method and Chebyshev rational approximation method (CRAM) are applied in the resonance, transport and burnup calculation to provide the macro cross-section library. In this paper, several critical experiments and burnup benchmarks, including B&W critical experiment, KRITZ critical experiment and Takahama-3 burnup benchmark, are applied in the validation of KYLIN-V2.0. KYLIN-V2.0 is validated by 21 cases in B&W-1484, 23 cases in B&W-1810 and 16 samples in Takahama-3. Eigenvalue, pin power distribution and nuclides density results of these cases show the good consistency compared with measured data. The KYLIN-V2.0 code shows good heterogeneous calculation ability for two-dimensional lattices.

Fission product depletion chain simplification method based on quantitative contribution function

Recently, an advanced neutronics lattice code called Kylin-2 has been developed by Nuclear Power Institute of China. In order to develop multi-group library for Kylin-2, a fission product depletion chain simplification method was proposed in this paper to generate burnup data for the multi-group library. This method constructed two quantitative contribution functions based on the thought of transmutation trajectory analysis (TTA), according to these contribution functions, a code was developed to identify important fission product nuclides and then a simplification depletion chain was reconstructed using these important nuclides. Verification and validation showed that this method successfully identifies important fission product nuclides and improve the depletion accuracy.

A new pin-resolved ultra-fine-group method based on global–local resonance treatment framework

In order to accurately consider intra-pin effects such as the spatial self-shielding effect and non-uniform temperature effect, several pin-resolved resonance self-shielding methods have been developed recently. The basic strategy of these pin-resolved resonance self-shielding methods is to divide resonance calculation into the whole domain calculation (global) and the 1-D fuel pin calculation (local), which we call the global–local resonance treatment framework. The spatial self-shielding effect, non-uniform temperature effect and resonance interference can be considered accurately by the 1-D fuel pin calculation with the ultra-fine-group method. Due to the simplified geometry, the computation time is short. In addition, the 1-D fuel pin calculation can be performed in parallel. This paper reviews pin-resolved resonance self-shielding methods that have been proposed and proposes a new pin-resolved ultra-fine-group method. These methods are implemented in the numerical nuclear reactor neutronics code SHARK developed by Nuclear Power Institute of China. Numerical results show good accuracy in the pressurized water reactor analysis.

Validation of PWR Neutronics Code Package TORCH V2.0 With Nuclear Power Plant Measurements

This article presents the verification and validation (V and V) of PWR neutronics code package TORCH V2.0 with nuclear power plant (NPP) measurements. The advanced nuclear power engineering design software, TORCH V2.0, was developed by the Nuclear Power Institute of China (NPIC), China National Nuclear Corporation (CNNC). Based on the two-step calculation scheme, TORCH V2.0 mainly contains lattice physics code for assembly homogenization, link calculation code for few-group constant parameterization, and core simulation code for few-group core calculation. The calculation modules of each code were already verified against various benchmark problems, whereas this article focuses on the V and V of linked code system. The measured values of the reactor startup physics test and NPP operation from six PWR NPPs (Daya Bay NPP, Ling Ao NPP, Fangjiashan NPP, Qinshan NPP, Hainan Changjiang NPP, and Fuqing NPP) were utilized to perform the comparison and analysis of V and V. Compared parameters of the reactor startup physics test include critical boron concentration, control rod integral value, boron differential value, and isothermal temperature coefficient. Compared parameters of the NPP operation contain critical boron concentration, assembly-wise power distribution, hot spot factor, and nuclear enthalpy rise factor. The results show that the software TORCH V2.0 has reliable calculation ability and can be applied in the PWR nuclear power engineering design which is based on square fuel assembly.

Validation of Doppler Temperature Coefficients and Assembly Power Distribution for the Lattice Code KYLIN V2.0.

This work is interested in verifying and analyzing the advanced neutronics lattice code KYLIN V2.0. Assembly calculations are an integral part of the two-step calculation for core design, and their accuracy directly affects the results of the core physics calculations. In this paper, we use the Doppler coefficient numerical benchmark problem and CPR1000 AFA-3G fuel assemblies to verify and analyze the advanced neutronics lattice code KYLIN V2.0 developed by the Nuclear Power Institute of China. The analysis results show that the Doppler coefficients calculated by KYLIN V2.0 are in good agreement with the results of other well-known nuclear engineering design software in the world; the power distributions of AFA-3G fuel assemblies are in good agreement with the results of the RMC calculations, it’s error distribution is in accordance with the normal distribution. It shows that KYLIN V2.0 has high calculation accuracy and meets the engineering design requirements.

The formation mechanisms of high burnup structure in UO2 fuel

The formation process of high burnup structure (HBS) in UO 2 fuel is accompanied by the subdivision of original grains and the formation of fission gas bubbles, which can accelerate the fission gas swelling and release, and correspondingly raises some safety concerns over the extension of operating life of the nuclear fuel in the reactor. The UO 2 fuel specimens were irradiated up to a maximum burnup of 120 GW·d/tU in High Flux Engineering Testing Reactor (HFETR) of Nuclear Power Institute of China (NPIC). High resolution scanning electron microscope (SEM) observations were carried out in the fuel specimens with different local burnups of 48–120 GW·d/tU. Our results show that there exist two restructuring mechanisms during the formation of the HBS in UO 2 fuel matrix: polygonization and recrystallization. Furthermore, a comparison with earlier experiments published at the CEA of France suggests that the deposition of fission fragments and recoil particles is the main mechanism responsible for the formation of small grains on the inner wall of large bubbles.

Test Analysis of Insulation and Moisture Resistance of Insulation Support Plate of Electrical Penetration in Containment of Nuclear Power Plant

The variation trend with temperature and relative humidity of water absorption, dielectric loss, relative permittivity, insulation resistance, volume resistivity, surface resistivity, and dielectric strength of polysulfone are measured. Polysulfone used as the insulating support plate on the medium voltage electrical penetration is designed and produced by Nuclear Power Institute of China (NPIC). The result shows that the relative humidity of environment has a significant effect on the water absorption of polysulfone. When the ambient temperature is 23℃ and the relative humidity increases from 30% to 98%, the corresponding water absorption increases from 0.012% to 0.106%, with a growth rate of 783.3%. The insulation resistance decreases gradually with the increasing of ambient temperature and relative humidity, with a maximum decreasing of 99.82%. However, the interphase insulation resistance of the insulation support plate of the electrical penetration is always greater than 200 TΩ. At the same time, the insulation resistance is greatly affected by its surface resistivity. The variation trend of insulation resistance and surface resistivity with ambient temperature and relative humidity is very close. The relative permittivity and dielectric strength are little affected by the ambient temperature and humidity. The insulation support plate of the medium voltage electrical penetration designed and produced by NPIC is with excellent moisture resistance and electrical insulation performance, and can work stably and reliably in high voltage, high temperature and high humidity environment.

Sensitivity Analysis of Multi-Layer Molten Pool Model of PWR Lower Head

The molten pool model of the lower head is an important model to evaluate the effectiveness of In Vessel Retention (IVR), which has been widely used in the safety evaluation of typical pressurized water reactor (PWR). The distribution and transfer process of the composition and heat of the melt in the pool of the traditional two-layer pool model and the three-layer pool model proposed in recent years are simulated, which is with the characteristics of complex relationship and strong nonlinearity. In order to provide a support for the optimization of the molten pool delamination model and the mitigation strategy of serious accidents, both global sensitivity analysis library (SALib) and IVR analysis code (CISER V2.0) developed by Nuclear Power Institute of China are used to analyze the sensitivity of four molten pool multilayer models and obtain the influence degree of the main input parameters on the key result parameters of each model. The sensitivity analysis results reflected a typical characteristics of the molten pool model. The radius of the lower head has a significant impact on the four molten pool multilayer models, while the input parameters that have significant influence on the key result parameters are basically the same in the Salay & Fichot model and the two-layer molten pool model, the initial mass of molten material has the greatest impact on the Esmaili model, and the density of the molten material has the greatest impact on the Seiler model.

Optimization of a Fuel Assembly for Supercritical Water-Cooled Reactor CSR1000

As one of the Generation IV nuclear reactors, the SCWR (supercritical water-cooled reactor) has high economy and safety margin, good mechanical properties for its high thermal efficiency, and simplified structure design. As the key component of nuclear reactor, the fuel assembly has always been the main issue for the design of the SCWR. The design of the fuel assembly for CSR1000 proposed by the Nuclear Power Institute of China (NPIC) has been optimized and presented in this study, which is composed of four subassemblies welded by four filler strips and guide thimbles arranged close together in the cross-shaped passage. Aiming at improving the hydraulic buffer performance of the cruciform control rod, the scram time and terminal velocity of control rod assembly were calculated to assess the scram performance based on the computational fluid dynamics and dynamic mesh method, and the mechanical property and neutronic performance of assemblies were also investigated. It has been demonstrated that the optimized fuel assembly had good feasibility and performance, which was a promising design for CSR1000.

Investigation on Rod Bundle CHF Mechanistic Model for DNB and DO Prediction Under Wide Parameter Range

For a water cooled reactor, the key thermal-hydraulic parameters span a wide range corresponding to different CHF regimes. Under accident conditions, due to the flow regime transition and interchannel mixing effect, the corresponding CHF can transition from the DO to DNB regime. In order to continuously and accurately predict DNB and DO regime CHF under wide parameter range for rod bundle channel, a comprehensive CHF mechanistic model covering the DNB and DO regime CHF prediction is established based on the rod bundle CHF-regime criterion. The DNB regime CHF mechanistic model of superheated liquid layer depletion under turbulence fluctuation bubbles and the mature DO regime CHF mechanistic model are combined to form the comprehensive CHF model. Furthermore, the comprehensive CHF model is assessed by 5 × 5 rod bundle CHF experimental data independently obtained by the Nuclear Power Institute of China (NPIC). The statistical evaluation and parametric trend analysis show that the maximum mean error of P/M is within ±22%, and the local pressure, mass flux, and quality do not have any effects on the average deviations of the predicted flux P from the measured flux M. This indicates that the comprehensive CHF mechanistic model can accurately and continuously predict the DNB and DO regime CHF in the rod bundle channel.

Description and verification of KYLIN-V2.0 lattice physics code

KYLIN-V2.0 is an advanced neutron transport lattice physics code developed by the Nuclear Power Institute of China (NPIC). This paper presents a brief introduction of the methodology used in the KYLIN-V2.0 code, and focuses on the verification of KYLIN-V2.0. In this paper, various fuel pin, fuel assembly models in the VERA, JAEA, and BEAVRS benchmark have been built by KYLIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3×3 fuel assembly case, and making a comparative analysis on the reference results given in benchmark, as well as the calculation results of the Monte Carlo codes RMC, MCNP, McCARD, etc. The numerical verification results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly. The effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worth of various fuel assemblies that are calculated by KYLIN-V2.0 are in good agreement with the reference.

Development and simple validation of the FAC_NPIC computer code for fatigue assessment

Fatigue is a significant degradation mode that affects nuclear power plants around the world.A fewself-designed computer codes have been developed forfatigue analysis. The Nuclear Power Institute of China has finalized an in-house computer code called FAC_NPIC for evaluating fatigue according to international standards. FAC_NPIC extracts stress data from finite element analysis results or manual input and is independent of specific finite element software. The “rubberband” peak-and-valley detection algorithm searches for extreme stress values automatically, and the modified rainflow-3D algorithm captures the features of secondary fluctuations. This paper presents transient combination schemes and how seismic loading effects are handled, as well asexamples of input and output files. A simple benchmark calculationwasconducted and compared with the literature. The results showed that all key features of FAC_NPIC work well.Despite differences in fatigue algorithms, good agreement with thestandards was achieved.