Thermal consideration of CANDU-SCWR sliding pressure startup through subchannel analysis

A startup system and startup procedures are designed based on the subchannel analysis for CANDU-SCWR sliding pressure startup. Lookup tables are selected to predict the CHF and PDO heat transfer due to their wide application range. Plant parameters are analyzed in detail. The results show that the maximum cladding surface temperature can be well restricted lower than criterion (850 °C), and the proposed startup procedure is feasible for CANDU-SCWR from the point of view of thermal-hydraulics.     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.     

FIDAS: Detailed subchannel analysis code based on the three-fluid and three-field model

The Film Dryout Analysis Code in Subchannels, FIDAS, has been developed with the main objective of predicting dryout and post-dryout heat transfer in a channel and in rod bundles. In FIDAS, two-phase flow consisting of continuous liquid film, continuous vapor and entrained droplets is modeled by a three-fluid, three-field representation of 12 field equations, i.e. three continuity, three energy and six momentum equations. FIDAS can predict dryout without any empirical CHF correlations by introducing annular flow modeling and the ‘film dryout criterion’. Experiments on film flow characteristics, subchannel flow and enthalpy distributions, dryout and post-dryout heat transfer in tubes and rod bundles were analyzed to demonstrate the performance of FIDAS. The predictions of FIDAS are in close agreement with the experiments.     

Subchannel analysis of CANDU-SCWR fuel

A cooperative study has been initiated at Xi'an Jiaotong University (XJTU) with Atomic Energy of Canada Limited (AECL) to develop a subchannel code ATHAS for preliminary analyses of flow and enthalpy distributions and cladding temperatures in CANDU fuel at super-critical water conditions. The code is applicable for transient and steady-state calculations. Then the paper uses the ATHAS code to analyze CANDU-SCWR which is operating at 25.0 MPa pressure. The results show that the maximum cladding-surface temperature of CANFLEX bundle is 804.1 °C, which is below the limit of design, and it is appropriate for use in the CANDU super-critical water-cooled reactor (SCWR) based on heat-transfer analysis.     

On the use of RELAP5-3D as a subchannel analysis code

The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries. Since no current subchannel codes can handle such fuel assembly designs, a new flexible RELAP5-based subchannel model was developed. It was shown that subchannel analysis of sodium-cooled fuel assemblies is indeed possible through the use of control variables in RELAP5. The subchannel model performance was then verified and validated in code-to-code and code-to-experiment analyses, respectively. First, the model was compared to the SUPERENERGY II code for solid fuel pins in a conventional hexagonal lattice. It was shown that the temperature predictions from the two codes agreed within 2% (<3.5 °C). Second, the model was applied to the Oak Ridge 19-pin test, and it was found that the measured outlet temperature distribution could be predicted with a maximum error of 8% (<7 °C). Furthermore, the use of semicircular ribs on the duct wall to flatten the temperature distribution in a traditional hexagonal assembly was explored by means of the newly developed RELAP5-3D subchannel model; the results are reported here as an example of the model capabilities.     

Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions.     

Multi-fluid multi-phase subchannel analysis of subassembly accidents of LMFRs

Analyses of experiments simulating hypothetical subassembly accidents such as a large-scale inlet blockage in a Liquid Metal Cooled Fast Reactor (LMFR) have been performed with computer program KAMUI. With the use of relatively simple but reasonable constitutive models, the code has been applied to the SCARABEE experiments BE+1 and APL1 to validate the analytical capability against the accident conditions under the multi-pin geometry. The results show that the key events such as sodium boiling, clad melting, molten clad relocation, molten clad freezing were adequately simulated taking into account the effect of heat loss to the coolant flow in the outside channel of the test section.     

Core and sub-channel analysis of SCWR with mixed spectrum core

The SCWR core concept SCWR-M is proposed based on a mixed spectrum and consists of a thermal zone and a fast zone. This core design combines the merits of both thermal and fast SCWR cores, and minimizes their shortcomings. In the thermal zone co-current flow mode is applied with an exit temperature slightly over the pseudo-critical point. The downward flow in the thermal fuel assembly will provide an effective cooling of the fuel rods. In the forthcoming fast zone, a sufficiently large negative coolant void reactivity coefficient and high conversion ratio can be achieved by the axial multi-layer arrangement of fuel rods. Due to the high coolant inlet temperature over the pseudo-critical point, the heat transfer deterioration phenomenon will be eliminated in this fast spectrum zone. And the low water density in the fast zone enables a hard neutron spectrum, also with a wide lattice structure, which minimizes the effect of non-uniformity of the circumferential heat transfer and reduces the cladding peak temperature.     

Thermo-hydraulic analysis for SCWR during power-raising phase of startup

The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR.     

Simulation of the distribution of flow and phases in vertical and horizontal bundles using the ASSERT-4 subchannel code

ASSERT-4 is a subchannel code based on the non-equilibrium equations of two-fluid flow. The paper briefly describes the equations and constitutive models used in the code, and reviews a number of validation exercises in which code results were compared to measurements in vertical and horizontal two-phase flows.     

Coupled thermal-hydraulics and neutron-physics analysis of SCWR with mixed spectrum core

The new SCWR conceptual design (SCWR-M) is proposed on the basis of a mixed spectrum core consisting of a thermal spectrum zone and a fast spectrum zone. This new core design is considered to be the hybrid of the existing thermal SCWR and fast SCWR cores. It combines the merits of both thermal and fast SCWR cores, at the same time minimizes their shortcomings. For the thermal zone, the difficulties in the mechanical design and the maximum cladding temperature can be reduced as far as possible by the co-current flow mode; and for the fast zone, a sufficiently large negative coolant void reactivity coefficient and breeding ratio can be achieved by the multi-layer arrangement of fuel rods.The performance, including the burn-up behavior, of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. The results obtained so far have shown that the mixed spectrum SCWR concept (SCWR-M) is feasible and promising.     

Sub-channel analysis of CANDU–SCWR and review of heat-transfer correlations

The paper presents the results of sub-channel analysis of CANDU–SCWR based on a wide review of heat transfer correlations. According to comparison with experiment data at different heat flux, Bishop Correlation is selected in SUBCHAN code to analyze CANDU–SCWR fuel channel. By detailed calculation of 43 fuel rods fuel channel in CANDU–SCWR, the paper gets the conclusion that the mass flux redistribution and reduction of heat-transfer coefficient at supercritical condition caused by the steep change of coolant density will limit the power of fuel channel in CANDU–SCWR.     

SCWR single channel stability analysis using a response matrix method

A system response matrix method, which directly solves the linearized differential equations in the matrix form without Laplace transformation, is introduced for the supercritical fluids flow instability analysis. The model is developed and applied to the single channel or parallel channel type instability analyses of a typical proposed Supercritical Water Reactor (SCWR) design. A uniform axial heat flux is assumed, and the dynamics of the fuel rods and water rods are not considered in this paper. The sensitivity of the decay ratio (DR) to the axial mesh size is analyzed and found that the DR is not sensitive to mesh size once sufficient number of axial nodes is applied. The sensitivity of the stability to inlet orifice coefficient is conducted for the hot channel and found that a higher inlet orifice coefficient will make the system more stable. The susceptibility of stability to operating parameters such as mass flow rate, power and system pressure is also performed. It is found that the SCWR stability sensitivity feature can be improved by carefully choosing the inlet orifice coefficients and operating parameters. The stability feature of the average channel is also analyzed with an equivalent inlet orifice coefficient. Finally, the manufacturing feasibility of the inlet orifices for both the hot channel and average channel is studied and found to be favorable.     

Development of a subchannel code for blockage accidents of LMFRs based on the 3D fuel rod model

To predict the thermal-hydraulic(T/H) parameters of the reactor core for liquid-metal-cooled fast reactors(LMFRs), especially under flow blockage accidents, we developed a subchannel code called KMC-FB.This code uses a time-dependent, four-equation, singlephase flow model together with a 3D heat conduction model for the fuel rods, which is solved by numerical methods based on the finite difference method with a staggered mesh. Owing to the local effect of the blockage on the flow field, low axia...     

CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential distribution of wall temperature and heat transfer, which is favorable for the design of SCWR fuel assemblies.     

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.     

The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis

In September 1988, the United States Nuclear Regulatory Commission issued a revised emergency core cooling system rule for light water reactors that allows, as an option, the use of best estimate plus uncertainty methods in safety analysis. To support the 1988 licensing revision, the United States Nuclear Regulatory Commission and its contractors developed the code scaling, applicability and uncertainty evaluation methodology to demonstrate the feasibility of the best estimate plus uncertainty approach. The phenomena identification and ranking table (PIRT) process, Step 3 in the code scaling, applicability and uncertainty methodology, was originally formulated to support the best estimate plus uncertainty licensing option. Through further development and application, the PIRT process has shown additional utility as a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. The generic PIRT process, including typical and common illustrations from prior applications that promoted further development of the process, are described. Analysis of the results of the prior applications is also described. The analysis results provide information that can help guide future applications of the process in a graded approach based on phenomena relative importance.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

Evaluation of conduction mixing lengths for subchannel analysis of finite LMFBR bundles

An analytical, two-dimensional, multi-region, multi-cell technique was developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for seven- and nineteen-rod bundles of different geometries and power distributions. The validity of the technique was verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential cladding temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect under practical conditions.A scheme was also developed to couple the two-dimensional distributed analysis and lumped parameter calculation such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a seven-rod bundle. The results of calculation were compared to those of three-dimensional analysis.