Maximum Cladding Surface Temperature Research Articles

Transient analysis of MTR research reactor during fast and slow loss of flow accident

Abstract The main objective for reactor safety is to keep the fuel in a safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accidents (DBAs), one of them is the loss of flow accident (LOFA), is required for assessing reactor safety. In this research, the safety aspects of 22 MW MTR research reactor under steady state and during loss of flow accident is studied. The flow transients considered include fast loss of flow accident (FLOFA) and slow loss of flow accident (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The analysis is done using PARET, a neutronics-hydrodynamics-heat transfer code. The transients were initiated from a full power with a flow trip point at 85% nominal. The calculated parameters are the temperatures of different components (fuel, clad and coolant) as a function of time for the hot channel. The results indicate that in both accidents the calculated maximum cladding surface temperature for the hottest channel of the reactor core does not exceed the allowable safety limit and the fuel integrity is maintained.

Three-dimensional transient analysis for control rods drop into CSR1000 core

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000MWe (CSR1000) core during the control rods (CR) drop into the core is analyzed and evaluated with the coupled three- dimensional neutronics and thermal-hydraulics SCWR transient analysis method. The 3-D CR drop transient analysis shows that while the high worth CR drop into the core, the core power in the initial stage of the CR drop process decreases rapidly. The core power descent rate exceeds the shutdown setting value, which will trigger the protection shutdown. While the low worth CR drop into the core, due to the reactivity feedback of the water density, the core power decreases slowly, and the core power descent rate cannot reach the shutdown setting value, which will not trigger the protection shutdown. In the transient process, the peak value of maximum cladding surface temperature (MCST) remains lower than the transient safety limit of 850 °C, whether it is a high worth CR drop or a low worth CR drop. The safety performance of CSR1000 core can be ensured in the process of CR drop into the core under the salient reactivity feedback of water density and the essential reactor protection system.

Research on Improved Simplified SCWR Assembly and Core Design

Abstract A new type of simplified assembly without “water rod” or solid moderator is identified to simplify China Supercritical Water-cooled Reactor with the rated electric power of 1000 MWel (CSR1000) core structure. A study on the properties of this new assembly has been carried out. Under the condition of heat full power, maximum cladding surface temperature (MCST) is 658 °C, very close to safety criterion 650 °C, and maximum linear heat generation rate (MLHGR) is 40.0 kW/m, slightly higher than safety criterion 39 kW/m. Considering rod stuck criterion, the maximum keff is 0.9833, lower than safety criterion 0.99. The discharge burn-up reduced to 21368 MWd/t, a sharp decreasing of 31.1%. This simplified assembly provides a new idea for solving the problem of complex structure of supercritical water-cooled reactor (SCWR).

Conceptual design of Super FR for MA transmutation with axially heterogeneous core

Supercritical water cooled Fast Reactor (Super FR) is featured with the large coolant density reduction by almost 1/10 from the core inlet to the outlet. Since Minor Actinides (MAs) not only have large neutron capture cross sections for thermal neutrons, but also can fission with fast neutrons, MA transmutation performance of Super FR may greatly depend on MA loading positions in the core axial direction of Super FR.However, such investigations have never been conducted. Hence, this study aims to design a Super FR transmutation core concept with the axial configuration of multiple layers of Mixed Oxide (MOX) and blanket fuels, with a focus on the influence of the large axial coolant density change on MA transmutation and core characteristics.With the design criteria of the negative Void Reactivity Coefficient (VRC), the Maximum Cladding Surface Temperature (MCST) < 650 ℃ and the Maximum Linear Heat Generation Rate (MLHGR) < 39 kW/m, three-dimensional neutronics and thermal-hydraulics coupled core burnup calculations have been carried out. Assembly-wise coolant flow rate distribution is determined to attain high core average outlet temperature and the core characteristics of different designs have been evaluated for the equilibrium core after the cores have reached equilibrium states with given fuel shuffling schemes. It has been shown that the MA transmutation amount is limited by deterioration of VRC due to increase of Pu enrichment for compensating the reactivity penalty by MA loading. Moreover, such influence has been found to be more significant in the lower region of the core, where the coolant density is relatively high. Hence, the core design with MA loading to the upper MOX layer is favorable for improving the MA transmutation performance. However, the trade-off relationship between the MA transmutation amount and thermal-hydraulics performance (increase of MLHGR and decrease of average outlet temperature) has been revealed. To overcome the issue, the core radial zoning has been applied and it has been found effective to suppress the trade-off relationship.

Study on thermal hydraulic characteristics under startup of SCWR

To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.

Accident analysis of supercritical water reactors during startup

To analyze the typical characteristics of accidents occurring during the startup of a supercritical water reactor (SCWR), a comprehensive SCWR system model was established with the SCTRAN analysis code, based on models of the high-performance light-water reactor (HPLWR) steam cycle, SCWR circulation startup loop, passive safety systems, and CSR1000 core parameters. According to the sequence of the loss of coolant flow accident (LOFA), without a safety system and trigger signal of the safety system under the rated condition, a new trigger signal during the startup was designed and then used to assess “LOFA,” “uncontrolled CR withdrawal accident,” and “loss of coolant accident (LOCA)” during the startup process. The results show that the new trigger signal can ensure the effective and timely response of the safety system, as well as reactor safety during the startup process. Moreover, the maximum cladding surface temperature (MCST) of the reactor peaks (850 °C, well below the safety criterion of 1260 °C) at the end of the fourth stage of the startup process in the case of LOCA.

Study on Startup Control Characteristics for CSR1000

The analysis of the startup system and startup characteristics are one important part of SCWRs design. In order to realize the analysis of the whole startup system, this paper proposes the new wall heat transfer model, based on the supercritical transient analysis code named SCTRAN. Then, the paper presents a control strategy for the startup process, including the control of coolant flow rate, core inlet temperature, system pressure, core thermal power, and steam drum water level. Different control schemes are set up according to different control objectives of the startup phases. Based on the CSR1000, an analytical model including the circulation loop and the once-through direct cycle loop is established, and four startup processes with control systems are put forward. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle loop meet the expectation, and the maximum cladding surface temperature does not exceed the limit temperature 650℃. The feasibility of the startup scheme and the security of the startup process are verified.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Startup Thermal Analysis of a Supercritical-Pressure Light Water-Cooled Reactor CSR1000

Supercritical-pressure light water-cooled reactors (SCWR) are the only water cooled reactor types in Generation IV nuclear energy systems. Startup systems, and their associated startup characteristic analyses, are important components of the SCWR design. To analyze the entire startup system, we propose a wall heat transfer model in a paper, based on the results from a supercritical transient analysis code named SCTRAN developed by Xi’an Jiao tong Tong University. In this work, we propose a new heat transfer mode selection process. Additionally, the most appropriate heat transfer coefficient selection method is chosen from existing state-of-the-art methods. Within the model development section of the work, we solve the problem of discontinuous heat transfer coefficients in the logic transformation step. When the pressure is greater than 19 Mpa, a look-up table method is used to obtain the heat transfer coefficients with the best prediction accuracy across the critical region. Then, we describe a control strategy for the startup process that includes a description of the control objects for coolant flow rate, heat-exchange outlet temperature, system pressure, core thermal power, steam drum water-level and the once-through direct cycle loop inlet temperature. Different control schemes are set-up according to different control objectives of the startup phases. Based on CSR1000 reactor, an analytical model, which includes a circulation loop and once-through direct cycle loop is established, and four startup processes, with control systems, are proposed. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle meets all expectations. The maximum cladding surface temperature remains below the limit temperature of 650℃. The feasibility of the startup scheme and the security of the startup process are verified.

Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 °C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 °C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking.

Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

CFD analysis of heat transfer in hexagonal subchannels of super fast reactor in upward flow

AbstractA super fast reactor is a fast spectrum, supercritical, water‐cooled reactor. This paper represents CFD analysis of heat transfer in hexagonal subchannels of super fast reactor using FLUENT in ANSYS. The numerical simulation of grid stability was done by considering different mesh sizes and the turbulence model for heat transfer of supercritical water was also carried out and compared with the experimental data. RNG k‐ϵ turbulence model with enhanced wall treatment was considered for simulations. Heat transfer and heat generation rate analysis of the outer surface rod wall is carried out with different subchannels by changing various parameters like boundary conditions and pitch‐to‐diameter ratio. The analyses reveal that the outer surface of the rod wall temperature decreases with increase in pitch‐to‐diameter ratio. Maximum coolant temperature rises in edge subchannels more than corner subchannels. Further analysis is carried out with different mass fluxes. Increases in mass flux has minimal effect on the maximum rod wall surface temperature. Maximum cladding surface temperature for the corner subchannel is less compared to the edge subchannel. Heat generation rate also decreases with increase in pitch‐to‐diameter ratio. This paper also investigates the buoyancy effect on subchannels with varying heat flux as boundary conditions considering constant mass flux.

Passive safety system of a super fast reactor

Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900°C. (2) Maximum cladding stress in circumferential direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be less than 1.0. Unfortunately some criteria could not be satisfied, such as the maximum fuel centerline temperature and maximum cladding stress. Therefore the influence of average linear heat generation rate, Pu enrichment and compressive stress to yield strength ratio are analyzed. Finally the improved fuel rod design of the new core is proposed.

Accidents and transients analyses of a super fast reactor with single flow pass core

The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST.

Subchannel analysis of a small ultra-long cycle fast reactor core

Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498°C on average and 551°C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498°C on average and 548°C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499°C on average and 538°C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

Numerical investigation on practicability of reducing MCST by using grid spacer in a tight rod bundle

The numerical investigation was carried out to reveal the practicability of reducing the maximum cladding surface temperature (MCST) within the inner sub-channel of a tight, hexagon rod bundle using commercial CFD code STAR CCM+ 6.04. The special heat transfer and pressure drop characteristics caused by four existing grid spacer designs were discussed in detail by analyzing the effects of grid strap length, different flow enhancing features and different Reynolds numbers. It was found that the local heat transfer within the grid strap is greatly enhanced due to the raised flow velocity. Both the standard grid spacer and the grid spacer with split-vanes cause decreased heat transfer in the downstream region. The friction drag is very influential in the tight rod bundle and can eliminate the positive effect of flow blockage on the heat transfer performance. The grid spacer with flow blockage discs induces relatively good heat transfer performance and higher pressure drop within sub-channels, indicating a tradeoff between the heat transfer augmentation and the pressure drop. The combination of multiple existing grid spacers can reduce the MCST to a certain level, but the corresponding disadvantages cannot be ignored. The improved grid spacer design was proposed based on the overall considerations of heat transfer and pressure drop characteristics and has been proved more suitable to widely reduce MCST for SCWR than any other grid spacer designs involved in present study.

Improvements of two-pass core design for super fast reactor

Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

Improved single pass core design for high temperature Super LWR

Core design of a supercritical-pressure light water cooled and moderated reactor (Super LWR) with single flow pass is carried out for achieving high outlet coolant temperature 500°C and simplifying upper core structure. Both the coolant in water rod and fuel channel flows up and is mixed at the upper plenum. The structure inside pressure vessel is simplified as a PWR by eliminating upper core moderator distribution/guide tubes applied in previous two pass core. The fuel assemblies loaded at peripheral core region are divided into four flow zones by using separation plates to adjust flow to the power for increasing average outlet temperature. The shuffling scheme is carried out separately for inner and peripheral core assemblies. Axial fuel enrichment is split into four segments to control the axial power peaking due to relatively large axial water density variation. Five fuel batches are adopted to increase the discharge burn-up. The equilibrium core is analyzed by employing neutronic/thermal-hydraulic coupled calculation. The numerical results show that all the design criteria are fulfilled by the maximum cladding surface temperature of 656°C with 500°C average core outlet temperature, maximum linear heat generation rate of 37.4kW/m and positive water density as well as shutdown margin of 1.45%dk/k.

Single-pass core design of a low-temperature Super LWR

A supercritical-pressure light water cooled and moderated reactor (Super LWR) with a single-pass flow scheme is developed for simplifying upper core structures. Both coolant in the fuel channels and the water rods flow upward and are mixed in the upper plenum. It eliminates the moderator guide/distribution tubes in the upper core that were used in the previous Super LWR design adopting two-pass coolant flow scheme. This core design adopts a four-batch fuel management scheme and an out–in fuel loading pattern. One hundred and twenty-one fuel assemblies with an active height of 3.7 m are included. The flow rate fraction for water rods is 3.5%, and the thermal insulator is used to keep the moderator temperature below pseudocritical temperature. The equilibrium core is analyzed by using neutronic and thermal-hydraulic coupled calculation. The results show that the maximum cladding surface temperature (MCST) is limited to 485 °C with the average outlet temperature of 400 °C. The inherent safety is fulfilled by the positive water density reactivity coefficient and sufficient shutdown margin. On the other hand, the investigation of average outlet coolant temperature varying with MCST is carried out to explore the maximum outlet temperature by employing current MCST criterion and single-pass core design. The average outlet temperature increases with the MCST, and it achieves 465 °C with the thermal efficiency of 43.1% at the MCST criterion of 650 °C. The structure inside the reactor pressure vessel is simplified as a pressurized water reactor.