Decay Heat Removal Capability Research Articles

Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available.

Thermal Hydraulic Safety Assessment of LLCB Test Blanket System in ITER Using Modified relap/scdapsim/mod4.0 Code

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code relap/scdapsim/mod4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.

Safe reactor depressurization windows for BWR Mark I Station Blackout accident management strategy

In order to evaluate the effectiveness of reactor depressurization within accident mitigation strategy and how to avoid core damage during Station Black-Out accident in a BWR Mark I plant, a GOTHIC model has been developed to support characterization of reactor safety systems performance. The GOTHIC model provides seamless coupled simulations of the reactor coolant system and the containment system. In this study, the time intervals (also called “safe reactor depressurization windows”) to initiate the reactor depressurization in order to optimize the early cooling strategy by injecting fire water and avoid clad failure are studied based on the decay heat removal capability of the reactor vessel coolant. This concept is instructive for the operation of the safety systems during the SBO accident mitigation. Sensitivity studies of several key parameters like reactor power, mass flow rates through RCIC system and fire water injection, and full open discharge coefficient of SRVs are performed to evaluate their impact on the safe reactor depressurization windows.

Research on capability of secondary passive residual decay heat removal system after Main Feedwater Line Break (MFLB) accident

All advanced NPPs are required to improve safety strategy to cope with design extension condition (DEC) after Fukushima accident. One design improvements for HPR1000 is adoption of secondary Passive Residual Heat Removal System (PRS) to cope with station blackout accident together with loss of all feedwater, which can help to strengthen defense in depth capability and enrich DEC mitigation measures. To make the best use of PRS decay heat removal capability, investigation of PRS startup performance and decay heat removal capability after MFLB accident are investigated based on HPR1000 RELAP5 model. PRS startup response featured by pressure, mass flow rate and void fraction are investigated under three startup strategies. Strategy 3(backflow valve open strategy) is recommended due to minimum fluctuation during initiating stage, in which strategy steam isolation valve is normally open and backflow isolation valve is used to initiate PRS. Then Main Feedwater Line Break (MFLB) accident was simulated with the assumption of loss of auxiliary feedwater. Response characteristics in primary loop and PRS are investigated. After PRS being initiated after scram with 60 s delay, natural circulation in PRS will establish rapidly, and decay heat will be removed under natural circulation operation mode. No overall boiling occurred in primary circuit before isolation of affected SG. Secondary PRS is adequate to remove decay heat after DEC.

Solving the Spent Nuclear Fuel Problem by Fissioning Transuranics in Subcritical Advanced Burner Reactors Driven by Tokamak Fusion Neutron Sources

The Georgia Tech concept of the Subcritical Advanced Burner Reactor (SABR) spent nuclear fuel (SNF) transmutation reactor and supporting analyses to date are summarized. SABR is based on the fast reactor physics and technology prototyped in Experimental Breeder Reactor-II (EBR-II) and proposed for the Integral Fast Reactor and the PRISM Reactor and on the tokamak fusion neutron source physics and technology that will be prototyped in ITER. Preliminary fuel cycle calculations indicate that subcritical operation would enable a proliferation-resistant fuel reprocessing cycle that would safely accommodate fuel with up to 100% TRU content and that introduction of SABRs in a 1-to-3 power ratio with light water reactors would reduce the required SNF high-level waste repository capacity (defined on the basis of decay heat released) by a factor of 10 to 100. Preliminary dynamic safety calculations indicate that SABRs could be shut down to the decay heat level by turning off the plasma heating power without core damage in loss of heat sink, loss of flow, and loss of power accidents, but that additional decay heat removal capability is needed in the case of total loss of primary or secondary system pumping power.

Microporous Coatings and Enhanced Critical Heat Flux for Downward Facing Boiling During Passive Emergency Reactor Cooling

During a reaction-initiated accident (RIA) or loss of coolant accident (LOCA), passive external-cooling of the reactor lower head is a viable approach for the in-vessel retention (IVR) of Corium; while this concept can certainly be applied to new constructions, it may also be viable for operational systems with existing cavities below the reactor. However, a boiling crisis will inevitably develop on the reactor lower head owing to the occurrence of critical heat flux (CHF) that could reduce the decay heat removal capability as the vapor phase impedes continuous boiling. Fortunately, this effect can be minimized for both new and existing reactors through the use of a cold-spray-delivered, microporous coating that facilitates the formation of vapor microjets from the reactor surface. The microporous coatings were created by first spraying a binary mixture with the sacrificial material then removed via etching. Subsequent quenching experiments on uncoated and coated hemispherical surfaces showed that local CHF values for the coated vessel were consistently higher relative to the bare surface. Moreover, it was observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit varied appreciably along the outer surface. Nevertheless, the results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward facing boiling by more than 88%. Moreover, the cold-spray process is amenable to coating the lower heads of operating reactors.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

Response characteristics of HPR1000 primary circuit under different working conditions of the atmospheric relief system after SBLOCA

To cope with SBLOCA in absence of High-Head Safety Injection (HHSI) from design of HPR1000, atmospheric relief system (originally named as VDA in French) is uniquely designed to help to trigger Middle Head Safety Injection (MHSI) or Low Head Safety Injection (LHSI) earlier through cooling primary system quickly after SBLOCA. To make the best use of VDA decay heat removal capability, primary and secondary system of HPR1000 was modeled with RELAP5/SCDAP computer code. After steady-state initialization, a cold leg 30mm break SBLOCA was simulated with six simulation conditions and five additional cases including availability of ACCU, different VDA discharge locations and area. Response characteristics of primary loop under different VDA working conditions are investigated. Pressurizer pressure decreases rapidly to lower level to trigger the reactor scram, VDA activation and accumulator safety injection sequently. Peak cladding temperature is 899.45K occurring at 222s, which is far below the safety limit. Activation of VDA can trigger ACCU SI earlier with a critical point, while positive reactivity will be introduced due to negative moderator temperature effect and Doppler effect. Larger VDA discharge capability will introduce larger reactivity feedback, as well as induce lower core level and SG level. It's suggested that VDA discharge condition should be chosen before the critical point, with the compromise with reactivity feedback introduced due to the negative moderator temperature effect.

Thermal Hydraulic and Safety Assessment of First Wall Helium Cooling System of a Generalized Test Blanket System in ITER Using RELAP/SCDAPSIM/MOD4.0 Code

The key objective of the test blanket module (TBM) program is to develop the design technology for DEMO and future power-producing fusion reactors. The proposed first wall of the test blanket system (TBS) is a generalized concept for testing in ITER, an experimental fusion reactor under construction in France presently. The first wall of TBM (TBM FW) directly faces the plasma and is cooled by the first wall helium cooling system (FWHCS), which is considered as a critical component from an ITER safety point of view. The scope of this work comprises thermal hydraulic analysis of the FWHCS of a generalized TBS and the assessment of postulated initiating events (PIEs) on the ITER safety with the help of thermal-hydraulic code RELAP/SCDAPSIM/MOD4.0. The three reference accidents: in-vacuum vessel (VV) loss of coolant accident (in-vessel LOCA), ex-vessel LOCA, and loss of flow accident (LOFA) in FWHCS are selected for the safety assessment. This safety assessment addresses safety concerns resulting from FWHCS component failure, such as VV pressurization, TBM FW temperature profile, pressurization of port cell (PC) and Tokomak cooling water system vault annex (TCWS-VA), and passive decay heat removal capability. The analysis shows that the critical parameters are under the design limit and have large safety margins, in the investigated accident scenarios. A comparative analysis is also carried out with the previous results to validate the results.

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. This paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.

Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30g thorium pebbles, for enhanced conversion.The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600°C.Results show that high conversion ratios (CR >0.96) and passive safety can be combined in a thorium PBR within a practical operating regime, which means a thermal power of 100MW or higher, 1000days total residence time of the breeder pebbles and fuel pebble handling times longer than 14.5s, like in the HTR-PM. With an increased U-233 content of the fresh driver pebbles (18w%), breeding (CR=1.0135) can already be achieved for a 220cm core and 80cm driver zone radius. While the decay heat removal is sufficient in this design, the temperature feedback of the undermoderated driver pebbles is too weak to compensate the reactivity insertion due to the xenon decay during a DLOFC without scram. With a lower U-233 content per driver pebble (10w%) it was found possible to combine breeding(CR=1.0036) and passive safety for a 300cm core and 100cm driver zone radius, but this does require more than a doubling of the pebble handling speed and a high reprocessing rate of the fuel pebbles. The maximum fuel temperature during a DLOFC without scram was simulated to be 1481°C for this design, still quite a bit below the TRISO failure temperature. The maximum reactivity insertion due to an ingress of water vapour is also limited with a value of +1497pcm.

B135 非常用復水器を用いた原子炉減圧・冷却システム(OS3 軽水炉・新型炉・原子力安全(3))

B135 非常用復水器を用いた原子炉減圧・冷却システム(OS3 軽水炉・新型炉・原子力安全(3))

Pool boiling CHF enhancement by graphene-oxide nanofluid under nuclear coolant chemical environments

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R&D efforts to use the reactor vessel with micro-porous coating and nanofluids boiling-induced coating. In present study, an experimental study has been conducted to investigate the viability of using graphene-oxide nanofluid under various coolant chemical environments to enhance CHF during ERVC. Pool boiling CHF experiments were carried out for the thin-wire heater with controlling the heater orientation from horizontal to vertical, or at 0<θ<90°. The dispersion stability of graphene-oxide nanofluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH), and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Finally integral effects of graphene-oxide nanosheets and chemicals on CHF limits were investigated. Results showed that graphene-oxide nanofluids were very stable under ERVC coolant chemical environments and enhanced CHF limits up to about 40% at minimum at 90° of angle (vertical orientation) and about 200% at maximum at 0° of angle (horizontal orientation) in comparison to pure water.

Development of a porous body model for decay heat removal studies in a pool type sodium cooled fast reactor

In this paper, the primary heat transport circuit of a pool type fast breeder reactor has been analyzed to demonstrate its decay heat removal capability without exceeding design safety limits even under complete power failure condition leading to natural convection flow in the system. A three dimensional model of large sodium pool (~12 m diameter) and complex inter-wrapper flow path through narrow gaps (3.4 mm) between a large number of heat generating subassemblies is simplified to a semi-3D porous body model which is computationally less intensive. The proposed semi-3D model has been validated against the detailed three dimensional model. The analysis has been carried out using the general purpose CFD code STAR-CD. The predicted results demonstrate that the decay heat removal system is conducive for natural convection and can remove the heat, keeping all the important temperatures within design safety limits.

C225 福島第1原子力発電所の非常用復水器の除熱能力評価(OS8 軽水炉・新型炉・原子力安全)

C225 福島第1原子力発電所の非常用復水器の除熱能力評価(OS8 軽水炉・新型炉・原子力安全)

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.

Flow transients performance in an MTR-type research reactor

During a flow coastdown event leading to slowing down coolant flow, the rate of heat removal from the fuel element must be sufficiently high to prevent meltdown. It is essential to estimate the flow rate change and the decay heat removal capability. In many studies complete pump operating characteristics are used in analytical solutions of the problem. Under the coastdown phenomenon, retarding torque replaces motor torque. In order to determine this torque, all the induction motor losses during the event are identified and where possible these loss parameters are measured. Stator and rotor core losses, stator and rotor stray load losses and magnetizing saturation and rotor conductor skin effects are taken into account. The basics equations for coolant flow and for the rotating parts of the centrifugal pump are subsequently derived for an MTR-type research reactor such Tehran Research Reactor (TRR). Then the equation of flow motion is solved with another one which predicts the pump speed during the coastdown transient. The results of the present work are validated by comparison with experimental and analytical studies of the similar work. The model shows good agreement with the present literature.

Innovations in the ENHS reactor design and fuel cycle

A number of approaches were explored for improving characteristics of the encapsulated nuclear heat source (ENHS) reactor and its fuel cycle, including: increasing the ENHS module power, power density and the specific power, making the core design insensitive to the actinides composition variation with number of fuel recycling and reducing the positive void coefficient of reactivity. Design innovations examined for power increase include intermediate heat exchanger (IHX) design optimization, riser diameter optimization, introducing a flow partition inside the riser, increasing the cooling time of the LWR discharged TRU, increasing the minor actinides' concentration in the loaded fuel and split-enrichment for power flattening. Another design innovation described utilizes a unique synergism between the use of MA and the design of reduced power ENHS cores. Also described is a radically different ENHS reactor concept that has a solid core from which heat pipes transport the fission power to a coolant circulating around the reflector. Promising features of this design concept include enhanced decay heat removal capability; no positive void reactivity coefficient; no direct contact between the fuel clad and the coolant; a core that is more robust for transportation; higher coolant temperature potentially offering higher energy conversion efficiency and hydrogen production capability.

Safety analysis of the US dual coolant liquid lead–lithium ITER test blanket module

The US is proposing a prototype of a dual coolant liquid lead–lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.