Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal–chemical experiment CS28-1, I – Steady state

To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H₂ at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought.     

Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal–hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal–hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.     

Core thermal–hydraulic design

The core thermal–hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively.As the results of the core thermal–hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis.On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal–hydraulic design gives conservative results.     

Development and validation of a CATHENA fuel channel model for a post-blowdown analysis of the high temperature thermal–chemical experiment CS28-1

To form a licensing basis for a new methodology for a fuel channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993].     

Treatment of the thermal–hydraulic uncertainties in the pressurized thermal shock analysis

The pressurized thermal shock (PTS) analysis is a quantitative analysis to calculate the vessel failure probability of the embrittled reactor pressure vessel. The PTS analysis consists of three major parts, such as the probabilistic safety analysis (PSA), the thermal–hydraulic analysis (T/H), and the probabilistic fracture mechanics (PFM) analysis. Because each analysis involves many parameters and assumptions associated with the uncertainties, it is important to identify and incorporate them into the analysis. Though the PSA and PFM analysis can be easily treated statistically, the thermal–hydraulic analysis results are very difficult to be treated statistically. Instead, sensitivity analyses of the thermal–hydraulic inputs were performed to understand the significance of the variation in the thermal–hydraulic inputs to the PFM analysis. In this study, the existing PFM code was modified to incorporate the uncertainties in the thermal–hydraulic inputs for the PFM analysis. The effects of the uncertainties in the thermal–hydraulic inputs for the vessel failure probabilities were evaluated using the modified code. The results showed the effects of uncertainties in the thermal–hydraulic inputs on the vessel failure probabilities are not significant for the ranges of the transient types. Even for the larger uncertainties, the effects on the vessel failure probabilities are small. Also, the effects of the thermal–hydraulic uncertainties vary depending on the transient characteristics such that the effects are greatest for the pressure dominant transient. Within the transient, the relative increases in the failure probabilities are greatest for the circumferentially oriented semi-elliptical flaws. It was found that the results of the sensitivity analysis using one standard deviation are conservative enough to bound the analysis results considering the uncertainties in the thermal–hydraulic inputs.     

CANDU reactor core simulations using fully coupled DRAGON and DONJON calculations

The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to predict the core behavior, fuel bundle burnups and local parameter information need to be tracked. The history-based approach has been developed to follow local parameter as well as history effect in CANDU reactors.     

CATHENA: A thermalhydraulic code for CANDU analysis

This paper describes the Canadian algorithm for thermal hydraulic network analysis (CATHENA) transient, thermalhydraulics code developed for the analysis of postulated upset conditions in CANDU®¹ reactors. The core of a CANDU reactor consists of a large number of horizontal pressure tubes containing fuel bundles. As a result of the unique design of the CANDU reactor, the CATHENA thermalhydraulic code has been developed with a number of unique modelling capabilities. The code uses a one-dimensional, two-fluid, nonequilibrium representation of two-phase flow. Some of the unique features of the CATHENA code are the one-step semi-implicit numerical method used and the solid heat transfer modelling capability that allows horizontal fuel bundles to be represented in detail. The code has been used in the design and analysis of CANDU-3, CANDU-6 and CANDU-9 reactors. The code has also been used for the design and analysis of the multiple applied lattice experimental (MAPLE) class of reactors and for the analysis of thermalhydraulic experimental programs conducted by Atomic Energy of Canada Limited (AECL).     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

钴调节棒在堆芯内的发热分析

秦山三期CANDU-6型重水堆中,为了生产工业和医用⁶⁰Co源,采用钴调节棒替换不锈钢调节棒。钴棒由于受到中子及γ射线照射产生热量,此外,⁵⁹Co被活化变成⁶⁰Co,而⁶⁰Co衰变放出的射线被调节棒自身吸收也会产生热量。因此,有必要研究钴调节棒的发热变化,为进一步分析钴调节棒的温度场及慢化剂的热负荷提供设计输入。本工作采用MCNP程序模拟秦山三期CANDU-6型重水堆的全堆芯（包括燃料、控制棒、调节棒、冷却剂和慢化剂等）几何结构,研究不锈钢调节棒和钴调节棒的发热率。将基于上述钴调节棒计算结果的最大发热率提供给热工进行稳态和事故分析,确保反应堆热工安全性。     

Development of nonequilibrium pressurizer model with noncondensable gas

To investigate thermal–hydraulic characteristics of a steam–gas pressurizer in the integral type reactor, the steam–gas pressurizer model based on the two-region nonequilibrium concept was developed and introduced into RETRAN-3D/INT code. The model includes an explicit solution method for the one-dimensional governing equations and the equation of the state solution method to determine the thermal–hydraulic state of the steam–gas pressurizer volume. In addition, the wall condensation model based on the diffusion layer modeling was included to consider the effect of the noncondensable gas. The developed model was verified with the results from the pressurizer insurge experiment conducted at Massachusetts Institute of Technology. From the verification results, it was concluded that the developed steam–gas pressurizer model can sufficiently predict the pressurizer transient and it can be used as a component model of the one-dimensional system code based on the homogeneous equilibrium model.     

REA analysis of the Iranian VVER-1000 core by modification of PRORIA code for annular pellets

The PRORIA code and its recent modifications are described here. The PRORIA code analyzes the transient response of the core against the reactivity increase caused by the control rod rapid withdrawal. The code solves and analyzes neutronic and thermal–hydraulic equations simultaneously. The code is designed for western PWR-type reactor performance. The equations representing thermal–hydraulic and neutronic should be modified to use the code to analyze VVER-1000 reactor core transients, because The VVER-1000 reactor fuel has a central hole in the fuel pellet. In a cylindrical solid fuel pellet, operation of an oxide fuel material at high temperature alters its morphology and the inner region is restructured to form a void at the center surrounded by a dense fuel region. Most of the restructuring occurs within the first few days of operation with slow changes afterward. Hence, the effects of a central hole in mathematical equations and in the transient are investigated. After the code modification, three accident scenarios with control rod ejection are simulated. The results are in good agreement with those reported in the plant’s FSAR. The results show that the peak fuel temperature in the hot fuel pin is lower than what the original code predicts by 150–500 °C. Furthermore, the Doppler reactivity effect, when the fuel pellet has a central hole, is higher than the solid fuel pellet.     

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together.Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary.The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

Reactivity insertion limits in a typical pool-type research reactor cooled by natural circulation

The course of reactivity insertion in a pool type research reactor, with scram disabled under natural circulation condition is numerically investigated. The analyses were performed by a coupled kinetic–thermal–hydraulic computer code developed specifically for this task. The 10-MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects. In general, it was found that the power, core mass flow rate and clad temperature under fully established natural circulation are higher for high-enriched fuel than for low enriched fuel. This is unlike the case of decay heat removal, where equal clad temperatures are reported for both fuels. The analysis of reactivity represented by the maximum insertion of positive reactivity ($0.73) demonstrated the high inherent safety features of MTR-type research reactor. Even in the case of total excess reactivity without scram, the high reactivity feedbacks of fuel and moderator temperatures limit the power excursion and avoid consequently escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation. The code can also be modified to provide an accurate capability for the analyses of research reactor transients under forced convection.     

CANDU-6 operation post-simulations using the reactor physics codes DRAGON/DONJON

The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to track the core behavior over a long period of operating history without having to use on-site measurement data, a consistent set of nuclear properties must be defined. The 2D as well as 3D capabilities of the DRAGON code are exploited to generate consistent two-group nuclear properties and increments using two different microscopic libraries. These properties are then used in a DONJON core-follow simulation of 220 full power days of operating history at the Gentilly-2 power plant. Comparisons with detectors show that differences tend to decrease with time. This core-follow application was pursued by post-simulations of reactivity mechanism measurements, which are shown to be in good agreement with reactor data. All these simulations demonstrate the DONJON capabilities of fuel management, detector reading evaluation and critical state determination.     

Thermal equivalence between nuclear fuel rod model and experimental heater simulator

Method of formulating nuclear fuel rod model (typical materials for pellet and clad, including usual gas-gap), which is thermally equivalent to multi-layer experimental heater simulator (any materials in contact, and/or without/with gas-gap) is presented. Thermally equivalent typical fuel rod model is defined. To validate it, the HECHAN model able to use different layouts of multi-layer rod/simulator geometry with constant and/or temperature dependent thermal–physical properties is used. Comparison with measured data for cladding surface temperatures in both pre- and post-CHF regimes proved applicability of this approach in transients. The rod model is easy to define, without changing the source of standard thermal–hydraulic code—only through input deck. It is qualified for further use in COBRA 3-CP and COBRA-TF codes for their validation on experiments using the particular heater simulator design.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.