Unprotected Loss Of Heat Sink Research Articles

Preliminary Analysis of Unprotected Loss of Heat Sink in Small Long Life Gas Cooled Fast Reactor

The failure of the secondary side in Gas Cooled Fast Reactor system, which may contain co-generation system, will cause loss of heat sink (LOHS) accident. In this study accident analysis of unprotected loss of heat sink due to the failure of the secondary cooling system has been investigated. The thermal hydraulic model include transient hot spot channel model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is also plays important role to limit the level of consequence during the accident. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate the failure of the secondary system by reducing power through reactivity feedback and remove the rest of heat through natural circulations based decay heat removal (RVACS system).

Unprotected Loss of Flow Accident in Small Long Life Gas Cooled Fast Reactor

In post Fukushima nuclear accidents inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Gas cooled fast reactors is one of the important candidate of 4th generation nuclear power plant and in this paper the safety analysis related to unprotected loss of flow in small long life gas cooled fast reactors has been performed. Accident analysis of unprotected loss of flow include coupled neutronic and thermal hydraulic analysis which include adiabatic model in nodal approach of time dependent multigroup diffusion equations. The thermal hydraulic model include transient model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is investigated. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations.

Transient analyses on loss of heat sink and overpower transient of natural circulation LBE-cooled fast reactor

Two transient analyses on a natural circulation LBE-cooled fast reactor were performed by using a Neutronics and Thermal-hydraulics Coupled Simulation Program NTC-2D. The two transients were unprotected loss of heat sink (ULOHS) and unprotected overpower transient (UTOP). For ULOHS, as the reactor is driven by natural circulation, the coolant mass flow rate decreased, but the simulation results showed that the reactor shutdown by itself and the peak temperatures of fuel and cladding were far smaller than the safety limits. For UTOP, the results showed that the coolant, cladding and fuel temperatures in the reactor increased and the more heat was taken away by the increase of the coolant mass flow rate. Finally, a new steady state was re-established by natural circulation.

Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

Preliminary transient analysis of the Autonomous Reactivity Control system for fast reactors

A preliminary parametric dynamic response study of the Autonomous Reactivity Control (ARC) system is performed for a large fast reactor core subjected to the following postulated design-extension accident scenarios: unprotected overpower (UTOP), unprotected loss of flow (ULOF) and unprotected loss of heat sink (ULOHS). The results show that the ARC system could prevent fuel and cladding failure as well as coolant boiling during UTOP and can lower transient temperatures during ULOHS. However, in case of ULOF, the design space examined the ARC system did not prevent exceeding permissible cladding temperature and introduced instability. A more thorough analysis is recommended.

Assessment of SFR reactor safety issues. Part I: Analysis of the unprotected ULOF, ULOHS and UTOP transients for the SFR(v2b-ST) reactor design and assessment of the efficiency of a passive safety system for prevention of severe accidents

In the frame of an AREVA – KIT cooperation in the years 2008–2013 on R&D issues associated with Innovative sodium cooled fast reactor concepts, several major topics were identified as being of mutual interest to both partners, one of which was the evaluation of severe transient behavior in Sodium-cooled Fast Reactors (SFRs) and the impact of newly conceived mitigation measures. The SFR design taken for the analysis was the SFR(v2b-ST) reactor design, and the system code to be used was selected to be the SIM-SFR code. The transients chosen for evaluation of the efficiency of mitigation measures were the unprotected loss-of-flow (ULOF), the unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP).

Code development and safety analyses for Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feed-water and lead bismuth eutectic (LBE) in the chimney of 3 m height, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have good passive safety. A new computer code is developed to investigate the thermal–hydraulic behaviors and safety performance of PBWFR in the present work. Unprotected rod run-out transient over power (UTOP) and unprotected loss of flow (ULOF)/unprotected loss of heat sink (ULOHS) are simulated to test and verify its safety. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal–physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more dangerous for PBWFR. It's suggested that it should appropriately reduce the maximum value of the control rods to mitigate the consequence of UTOP due to good reactivity feedbacks in the core.

Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

The molten salt reactor (MSR) is one of the Generation IV reactors. The fuel is dissolved in the carrier salt and circulates in the loop. The technologies are different from that in the solid-fuel reactors. In this work, the attention is focused on development of safety analysis tool for the MSRs. A single channel model, a heat transfer model, a heat sink model and a liquid-fuel point kinetic model that takes into account the effect of circulation are employed. The validation of this code is done with the experimental data of pump coastdown and startup in MSRE. The unprotected loss of heat sink (ULOHS), combination of ULOHS and unprotected loss of flow (ULOF) are performed on the Molten Salt Actinide Recycler and Transmuter (MOSART). This work aims to study temperature fluctuation, corresponding power change, effect of flow delayed neutron precursors and temperature reactivity feedback in transient accident to examine the inherent safety design of MOSART. The transient results reveal that the large negative temperature feedback coefficients guarantee MOSART inherent safety and the range of temperature is within the safety margin in case of combination of accidental events.

Development of technical basis in the initiating and transition phases of unprotected events for Level-2 PSA methodology in sodium-cooled fast reactors

Abstract A probabilistic safety assessment (Level-2 PSA) methodology was developed for comprehensive risk evaluation of sodium-cooled fast reactors. As part of this development, in this paper, phenomenological event trees were developed as well as technical database to quantify the probability of event sequences in the Level-2 PSA, focusing on the initiating and transition phases of unprotected events. Typical and important accident categories were selected: unprotected loss of flow (ULOF), unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOHS). Based on the state-of-the-art knowledge, the headings of these event trees were selected so that dominant factors in accident consequences can be represented appropriately. For each of the headings, available information for the probability quantification were reviewed and integrated as the technical database for the Level-2 PSA. It was clarified that the headings of the ULOF category, for which experimental database and evaluation models have been reasonably established, can be commonly applied to certain part of the different accident categories except for some specific points, which were identified in this study. For the ULOHS category, an additional event tree is necessary before the core disruption providing various boundary conditions for the initiating phase. For the transition phase, dominant factors were also identified through parametric analyses. In the Japan sodium-cooled fast reactor, an inner duct is introduced into a fuel subassembly for enhancing molten fuel discharge from disrupted core in the transition phase. The parametric study showed that the analytical case without the fuel discharge through the inner duct resulted in an occurrence of recriticality regardless of the fuel discharge through control-rod guide tubes. This suggests that the fuel discharge through the inner duct is essential to avoid severe recriticality in the transition phase.

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead–bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal–hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320–420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.

An Enhanced Code for the Safety Analysis of Pool-Type Sodium-Cooled Fast Reactors

The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.

Safety analysis of small long life CANDLE fast reactor

Safety analysis for small long life fast CANDLE reactor was performed with ULOF (unprotected loss of flow), SDRW (unprotected shut down rods withdrawal), ULOHS (unprotected loss of heat sink) and LB (local blockage) accidents. The employed reactor system is based on the former steady state research. The core with 1.0 m radius and 2.0 m length produces 200 MW thermal power in steady state, using enriched N-15 natural uranium as fresh fuel and lead bismuth as coolant. The former 3 accidents were simulated without scram by neutronic-thermal hydraulic calculation coupled with stationary diffusion calculation. The LB accident was simulated by transient thermal hydraulic calculation only, because in this accident the neutronic factors basically do not change. The analysis results show that the proposed small CANDLE fast reactor can survive all the accidents without any active protection.

Model development for analysis of the Korea advanced liquid metal reactor

The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.

1923 LBE 冷却自然循環高速炉の崩壊熱除去特性

This paper describes numerical analysis for a Lead-Bismuth eutectic (LBE) cooled natural circulation fast reactor. A transient analysis for the un-protected loss of heat sink (ULOHS) condition has been performed by the combined multi-dimensional plant analysis code (MSG-COPD). As a result, thermal-stratification phenomenon has been observed in the lower-plenum of the reactor vessel during the transient. It has become clear that the thermal-stratification should be prevented in a LBE-cooled natural circulation reactor's design.

Feasibility study of nitride fuel core and recycle system toward self-consistent nuclear energy system

A concept of nitride core and recycling system in relation to SCNES (1) has been investigated and it has been found that the safety of intact core is enhanced by zero burnup reactivity and negative feedback at ULOF (Unprotected Loss of Flow) and ULOHS (Unprotected Loss of Heat Sink). The self-actuating modules are examined so as to eliminate recriticality of degraded core. The nitride core can employ transuranics-added fuels and burn long life radioactive fission products after isotope-separation of FPs. The enrichment of 15N and the laser isotope separation seem to need about 2.4% of produced electric energy in the preliminary evaluation.

A private venture. The Mescalero/utility storage initiative progresses

A preliminary parametric dynamic response study of the Autonomous Reactivity Control (ARC) system is performed for a large fast reactor core subjected to the following postulated design-extension accident scenarios: unprotected overpower (UTOP), unprotected loss of flow (ULOF) and unprotected loss of heat sink (ULOHS). The results show that the ARC system could prevent fuel and cladding failure as well as coolant boiling during UTOP and can lower transient temperatures during ULOHS. However, in case of ULOF, the design space examined the ARC system did not prevent exceeding permissible cladding temperature and introduced instability. A more thorough analysis is recommended.

Freeing the nuclear industry

In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

Accident analysis of lead or lead—bismuth cooled small safe long-life fast reactor using metallic or nitride fuel

Safety analysis of a lead or lead—bismuth cooled small safe long-life fast reactor was performed. It is proposed that the reactor be used in relatively isolated areas, and operated to the end of its life without refueling or fuel shuffling. In the present paper the reactor power and lifetime are set at 150 MWt and 12 years respectively. In order to assume its safety performance, the following accidents without scram were simulated with neutronic-thermal-hydraulic analysis: unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), simultaneous ULOF and UTOP accidents, and simultaneous ULOF, UTOP and unprotected loss of heat sink (ULOHS) accidents. For each type of accident, four types of long-life small reactor (lead cooled metallic fueled, lead cooled nitride fueled, lead-bismuth cooled metallic fueled, and lead—bismuth cooled nitride fueled) were analyzed. It is shown that all the proposed designs can survive these accidents without requiring help from the operator or active devices.

Passive Safety Features of a Bottom-Supported Fast Breeder Reactor Vessel

A system dynamics analysis is applied to a pool-type fast breeder reactor to examine the influence of a bottom-supported reactor vessel (BSRV) design on anticipated transient without scram (ATWS) events such as an unprotected loss of flow (ULOF), an unprotected loss of heat sink (ULOHS), and an unprotected transient overpower (UTOP) by using the ARGO safety analysis code. The BSRV enhances negative feedback because of the differential displacement between the core and the control rod as compared with a top-supported reactor vessel. In particular, the BSRV has the potential, especially in a mixed-oxide-fueled core, to mitigate the design requirements to prevent boiling of the coolant during an ULOF and ULOHS through the elongation of the primary flow coastdown and enhancement of the axial expansion of the control rod drive line. In the metallic-fueled core, the effects of the BSRV on the ATWS events are diminished by the limitation of the sodium temperature increase.