Loss Of Heat Sink Accident Research Articles

Thermal-hydraulic characteristics analysis of unprotected accident and protection control strategy for helium-xenon cooled reactor system

The small modular helium-xenon cooled reactor has the advantages of simple and compact system structure, short construction cycle, light weight and small volume, which is very suitable for energy supply in remote area and small nuclear propulsion device. The research on accident characteristics and protection control strategies is very important to ensure the safe and reliable operation of the reactor system and mitigate the consequences of the accident. Therefore, the safety characteristics analysis program of helium-xenon cooled reactor system is developed based on Modelica language, and the transient verification of typical working conditions is carried out. The maximum relative error of the verification results was 7.91 %. Through the program, the unprotected transient conditions and corresponding protection control strategy transient conditions of external load loss accident, partial loss of heat sink accident, main valve accidental closing accident and Brayton unit efficiency step decline accident are simulated. The results show that the loss of all external loads will lead to rotation speed overspeed. The introduction of negative reactivity or bypass control can effectively control the rotation speed rise, but the introduction of negative reactivity is better than bypass control. Under unprotected control, the partial loss of the heat sink, the accidental closing of the main valve and the step down of the Brayton unit efficiency will all lead to the rapid decline of the rotation speed and mass flow rate, and there is a risk of the reactor temperature exceeding the limit. The rotation speed and mass flow rate can be effectively stabilized by load following to prevent the temperature exceeding the limit.

Preliminary analysis of maximum hypothetical accident for a solid core space nuclear reactor power system

In order to study the safety characteristics of the solid core space nuclear reactor power (SNRP) system under the maximum hypothetical accident, a two-dimensional entire core transient heat transfer analysis model was established, and the key parameters response characteristics of the ultra-small lithium-cooled SNRP under two maximum hypothetical accidents, namely, loss of heat sink (LOHS) and loss of coolant accident (LOCA), were calculated and analyzed. In the maximum hypothetical accidents, the coolant cooling capacity is lost, thus the core decay power is discharged into space only through the radiation of the residual heat removal system and the side surface of the reactor vessel. During the accidents, the heat transfer of the core deteriorated, and the fuel temperature may rise to the melting point, resulting in radioactive leakage. The results show that: (1) In the LOHS accident, the maximum fuel temperature reaches 2150 K at 550 s, and the pressure of the primary system volume accumulator continues to increase to the set system pressure safety limit of 2 MPa, resulting the primary loop overpressure failure. And the fuel matrix temperature is close to the set cladding limit temperature of 2200 K; (2) In the LOCA, the deterioration of heat transfer in the core makes the temperature increase rapidly, reaching a maximum of 3016 K at about 630 s, which is very close to the melting temperature of UN fuel 3123 K. As the decay power decreases, the maximum core temperature decreases to less than 1600 K after 24 h of the accident. The auxiliary cooling system of the solid core SNRP system under the maximum hypothetical accident is optimized, and the design parameters of the auxiliary cooling system meeting the safety requirements are obtained.

Conceptualization of the micro research reactor cooled by heat pipes (MRR-HP), part-II: safety analyses

ABSTRACT The safety of the novel Micro Research Reactor cooled by Heat Pipes (MRR-HP) design is explored. This is achieved by developing a 3D thermal conduction model using STARCCM+ software, besides, coupling it with the 3D neutronics model for reactor dynamics. The MRR-HP has a unique design characteristic, which is the implementation of dual-sets of heat pipes (HPs) in the monolithic-block: Water-HPs for operation, and Potassium-HPs for accident/post-accident scenarios. Water-HPs cool the reactor effectively, allowing for maximum fuel temperature (K). However, water-HPs may suffer from: Loss-of-heat-Sink-Accident(LOSA), which may occur due to blackout, since it requires forced convective cooling in the condenser. Non-Condensable-Gases-Generation-Accident (NCGGA), due to radiolysis. These events cause significant melting in the monolithic-block and fuel rods after 3 and 6 hours, respectively. This paper proves the reliability of potassium-HPs as a passive countermeasure against LOSA and NCGGA, while natural convective cooling in the condenser is applied. Cold-state no-scram Reactivity-Insertion-Accident (RIA) scenarios with an insertion of 2.35$ are investigated, showing that no partial or significant melting occurs, due to the negative reactivity feedback of the MRR-HP. Besides, Potassium-HPs are essential in removing the decay heat, for warm-state RIA, after water-HPs dry out. This paper proves the robustness of the MRR-HP and encourages its utilization for remote operation.

Inherent Safety Characteristics of Lead Bismuth Eutectic-Cooled Accelerator Driven Subcritical Systems

An accelerator driven subcritical system (ADS) is a new nuclear energy system which could not only produce clean energy but also incinerate nuclear waste. In this paper, inherent safety analysis of an ADS is performed with neutronics and thermal-hydraulics coupled code named ARTAP. Five typical accidents are carried out, including the cases of proton beam interruption, transient overpower, reactivity insertion, loss of flow, and loss of heat sink. The transient simulations are performed in the average channel and the hottest channel of the fuel pin in the ADS core. The simulation results for beam interruption show that the highest temperature of the pellet is in the middle of the fuel element in the average channel, while the peak temperature of the cladding is in the top of the fuel element. After the beam is interrupted for 20s, the maximum temperature drops at the fuel center, the cladding inner surface, and the outlet coolant in the hottest channel are 644.46K, 162.27K, and 136.42K respectively. For transient overpower accidents with the increase of beam intensity, the maximum temperature of the fuel and the cladding are below the safety limit. Concerning the reactivity insertion accident, it is found that the ADS has good inherent safety and its margin of criticality safety is large. The calculation results for loss of flow show that the power drop is small due to low sensitivity of the subcritical core to negative reactivity feedback, and the maximum temperature of the cladding reaches 1726K, which means the fuel element would rupture. However, the power and the temperatures of fuel, cladding, and coolant could decrease quickly to the safety level after the proton accelerator is cut off under a loss of flow accident. The results also show that the peak temperature of the cladding is lower than the safety limit under a loss of heat sink accident. The present simulation results reveal that the ADS has a remarkable advantage against severe accidents. It also implies that its inherent safety characteristics could ensure reactor shutdown by cutting off the proton beam after accidents occur.

Heat transfer analysis of heat pipe cooled device with thermoelectric generator for nuclear power application

Heat pipe cooled reactor featured with compact structure, high reliability, and inherent safety could be applied for the energy system for multipurpose. Numerical simulation methods are widely used to simulate the normal operating conditions of heat pipe reactors and even for accident conditions. In this paper, the ANSYS Fluent is used to perform thermal calculations on the heat pipe thermoelectric generator (HPTEG) prototype for both normal and accident conditions which include heat pipe failure and loss of heat sink accident conditions. The results show that the numerical simulation results are in good agreement with the experimental results, which proves that the fidelity of heat pipe model. Further analyses show that the failure of heat pipe would cause an increase in the power and operating temperature of the adjacent heat pipes, which may increase the possibility of failure of adjacent heat pipes. The maximum core temperature of the HPTEG prototype under the loss of heat sink accident is 1600 K, which exceeds the copper’s melting point temperature 1356 K. It is obvious that the loss of heat sink accident would cause the core temperature increase sharply, which is likely to cause damage to the core. This work provides reference to safety analysis and evaluation of the heat pipe cooled reactor.

Operation and safety analysis of space lithium-cooled fast nuclear reactor

As the power requirements for deep space exploration continue to increase, space nuclear power systems are imperative. In this paper, a set of models are established for the lithium-cooled space reactor combined with the Stirling engine. The space lithium-cooled reactor system is modeled, and the result of the steady-state is checked with the maximum relative error of 13.3%. Moreover, the characteristics under the unprotected reactivity insertion accident (URIA) and the unprotected loss of heat sink accident (LOHA) are obtained and analyzed. The results showed that: a) the solar heat flux cause the radiator temperature to fluctuate but has limited impact, b) under LOHA, the temperature of hot spot decreases rapidly from 1436 K to 1423 K, verifying the inherent safety of the system, c) under URIA, the hot spot temperature rises to 1546 K within 160 s. This work provides a solid basis for the design and analysis of space nuclear systems.

Evaluation of a freeze-tolerant decay heat removal system redundancy for fluoride salt-cooled high-temperature reactors (FHR)

Typical FHRs use a direct reactor auxiliary cooling system (DRACS) for decay heat removal under a variety of accident scenarios where active normal shutdown cooling is unavailable. Response surface method has been used for evaluating the reliability of the passive system. In this paper, a loss of heat sink (LOHS) accident sequence is evaluated using the response surface methodology. Response surfaces on maximum core outlet temperature and DRACS freezing time has been presented for evaluating the effects of friction loss, heat transfer coefficients and partial and full blockages of the DRACS. The present study shows that for one DRACS failure situations, the system is unconditional safe. But, for two DRACSs failure situations, there is probability that peak core outlet temperature (PCOT) will not exceed the allowable temperature. During the LOHS accident, the DRACS would not freeze within 3.33 days. This study will provide an insight for investigating DRACS redundancy and functional failure.

Dynamic simulation of the gas-cooled space nuclear reactor system using SIMCODE

One dimensional space nuclear reactor system analysis code—SIMCODE is developed and five typical transient operation conditions are calculated, including the system startup, the loss of flow accident, the loss of heat sink accident, the reactivity insertion accident, and the small loss of coolant accident. These conditions are part of the key performance of the space nuclear reactor system and are related with both the individual characteristics of each component and the coupled characteristics between all the components. By analyzing the simulation results, system transient performance and coupled system characteristics are acquired. The information can provide helpful information for the system evaluation, safety system design, and emergency strategy setup. Based on the simulation, system control suggestions are also proposed and will be beneficial for further researches.

Experimental investigation on instability characteristics of loss of heat sink accident in a natural circulation system

Loss of heat sink (LOHS) accident in a natural circulation system has a complex event sequence scenario. Flow instability phenomena in LOHS accident involves multiple flow patterns and variety instability mechanisms. Through experimental investigation, a methodology analyzing instability under LOHS is established. It is concluded while slug flow occurring in heating channel, the system falls into geyser instability region which is proven as pressure drop oscillations (PDO) accompanied by wave mode changing. While annular flow occurs, natural circulation system forms stable two-phase flow transitorily and losses stability again. The post-annular-flow instability cannot form a regular vibration wave mode, and annular flow absorbs flowrate oscillation instead of pressurizer. Finally, average flowrate approaches to 0 and critical heat flux (CHF) happens.

Analysis of unprotected loss of heat sink accident in compact high temperature reactor cooled under natural circulation of LBE coolant

India is developing 100kWth Compact High Temperature Reactor (CHTR) to demonstrate the technology enabling high-temperature process heat applications such as hydrogen production. The reactor core is envisaged with 233U-Th based TRISO coated fuel particles and beryllium oxide moderator in prismatic fuel assemblies for high-temperature operation. The core heat is designed to be removed by lead–bismuth eutectic (LBE) coolant under natural circulation in vertical channels of CHTR at nominal power. The heat generated in the primary circuit is to be transferred to the secondary side using high-temperature sodium heat pipes. The design and safety of the prototype core necessitate comprehensive, integrated neutronics-thermal hydraulics analyses of anticipated transients. In this context, the Unprotected Loss of Heat Sink Accident (ULOHSA) at full power during the beginning of the core life cycle is investigated with indigenous 3D multi-physics code ARCH-TH. The key parameters defining neutronics and thermal-hydraulics of the core have been assessed in the steady-state as well as during the transient. The findings, along with the model of the simulation and the code validation are discussed in this paper. The analyses indicate that the core is being shut down passively due to reactivity feedbacks and the loss of heat sink leads to reverse flow of the coolant in the low powered channels during the transient. However, the coolant and fuel peak temperatures are observed to stay well below the leak-tightness limit of TRISO particles and boiling point of LBE, even considering the failure of shutdown systems. The modifications in the design of CHTR are also investigated to pacify the condition of flow reversal during such transients.

Performance analysis of the SPRHRS for the 1000 MWe class PWR in coping with station black out

In order to improve the safety and reliability of the Chinese nuclear power plants (NPP), an innovative passive nuclear safety design, the secondary passive residual heat removal system (SPRHRS) is proposed for the light water reactors in China. Previous research work has shown that under design-basis accidents, SPRHRS can mitigate the results by partially or completely replacing the traditional emergency water cooling system. However, the Fukushima Daiichi nuclear power plant accident implies necessity of further assessments of NPPs capability to withstand threats resulting from accidents beyond design-basis accidents. In this study, the nodalization of a 1000 MW class PWR with the SPRHRS design was conducted using RELAP5/MOD3.4. Responses of the PWR under the loss of heat sink accident caused by the station blackout (SBO) was analyzed. The results showed that the residual heat can be effectively removed by the SPRHRS under SBO. In addition, sensitivity studies about the effects of the coolant tank volume and the isolation valve opening condition of the SPRHRS on its residual heat removal capability were also performed. The results indicate that these two main design elements have significant influence on the operation characteristics of SPRHRS. For the SPRHRS design in this study, the appropriate value of the coolant tank volume is 400 m3 and the appropriate valve opening signal is set to be triggered when the SG shell side collapsed liquid level drops lower than 20% of its full value. The present work would be instructive for the design of passive safety systems of Chinese NPPs.

Investigations of accelerator reliability and decay heat removal for accelerator-driven system

To realize the feasible accelerator-driven system (ADS) concept, the investigations for the reliable accelerator and conceptual plant design considering safety issues were performed. As the reliable accelerator concept, the double-accelerator concept was proposed to reduce the beam-trip frequency. The estimated beam-trip frequency with the double-accelerator concept using the J-PARC LINAC operation data showed that the beam-trip frequency was significantly improved with the comparison of the single accelerator result.The basic investigation of the primary reactor auxiliary coolant system (PRACS) was performed for the safety design of the LBE cooled ADS. The concept which the PRACS heat exchanger was integrated to the steam generator was proposed and the transient analysis in the loss of heat sink accident was carried out. The result presented that the decay heat removal was appropriate when the operation of the PRACS succeeded. The feasibility of the ADS in the beam trip accident was indicated through this study.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

Flow regime transition criteria for co-current downward two-phase flow

Downward two-phase flow is observed in light water reactor accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) due to loss of feed water or a secondary pipe break, and so, it is vital to have a thorough understanding of the flow mechanisms and regimes. With this point of view, flow regime transition criteria for vertical downward flow for a range of pipe diameters of 24–101.6 mm has been developed. Several models looked at the radial distribution of the bubbles and the wake effect of leading bubbles while others looked into the Kelvin-Helmholtz instability seen at the gas-liquid interface. The newly developed criteria have been compared to flow regime maps obtained via subjective and objective means, consisting of air-water data at atmospheric conditions as well as at an elevated pressure of 0.2 MPa. The new model is also compared to flow regime maps developed with different inlet conditions. Overall, the present model showed good agreements with the available data, with the exception of several 50.8 mm ID flow regime maps of different inlet conditions as well as a self-organizing neural network. This study also highlights the need for a more objective and consistent flow regime map data for large diameter pipes, the identification of cap-bubbly and churn-turbulent flows in these maps, and the deviations observed between a supervised and self-organizing neural network (SONN).

Experimental study on accident transients and flow instabilities in a PWR-type small modular reactor

Experimental study on natural circulation flow instabilities is of great importance for the safety analysis in a PWR-type SMR, especially for accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS). In this study, an experimental natural circulation facility was built by scaling down from a typical PWR-type SMR. The scaling ratios were derived from non-dimensional field and constitutive equations of the drift flux model. The test facility has a height of 3.44 m with an operating pressure limit of 1.0 MPa. Two kinds of tests, the blowdown test and cold-blowdown test were performed. The blowdown test was designed to simulate the low pressure phase (<1 MPa) of a LOCA event and evaluate the performance of the emergency core cooling system (ECCS). The test results indicate that the ECCS can keep the RPV water level above the core, and the RPV pressure keeps decreasing during the entire accident transients. The cold-blowdown test was designed to simulate the whole transients of the system blowdown at low pressure conditions. The oscillations of the pressure drop, natural circulation rate, void fraction and inlet temperature were observed and analyzed. The cold-blowdown test results were compared with the RELAP5 simulation. Although some discrepancies existed at the initial blowdown phase, the code calculation agreed well with the experiment data.

Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient.

Current status of accident analysis for Korean HCCR TBS

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Design and transient analyses of emergency passive residual heat removal system of CPR1000

The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the steam impact on the EPRHRs, and improve the stability of the system.

Design and transient analyses of emergency passive residual heat removal system of CPR1000. Part Ⅰ: Air cooling condition

The steam generator secondary emergency passive residual heat removal system (EPRHRS) is a novel design for the conventional generation Ⅱ+ reactor CPR1000. The EPRHRS is designed to improve the safety and reliability of CPR1000 by completely or partially replacing the traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The EPRHRS consists of a steam generator (SG), a heat exchanger (HX), an air cooling tower, an emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, a model of the primary loop system and the EPRHRS was developed using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system affected by the EPRHRS. The transient characteristics of the primary loop system and the EPRHRS were calculated in the event of the feed line break accident. Sensitivity studies were also conducted to investigate effects of the main parameters of the EPRHRS on the transient characteristics of the primary loop and the EPRHRS. The EPRHRS could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRS could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and the EPRHRS loop, respectively. The present work is instructive for engineering design of the EPRHRS for Chinese NPPs.

ICONE19-43763 DESIGN AND TRANSIENT ANALYSES OF PASSIVE EMERGENCY FEEDWATER SYSTEM OF CPR1000 : PART I: AIR COOLING CONDITION

The steam generator secondary passive emergency feedwater system is a new design for traditional generation II+reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively.