Boiling Water Reactor Research Articles

Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

Abstract The economic simplified boiling water reactor (ESBWR) is a boiling water nuclear reactor of generation III+. The U.S. Nuclear Regulatory Commission (NRC) approved the ESBWR design as the world's best light-water nuclear reactor in 2014. It has the lowest core damage frequency (industry standard indicator of safety) of any Generation III or III+ reactor. It can cool automatically for more than seven days without using electricity or human intervention. During the operation, the ESBWR is designed to produce electricity while emitting almost no greenhouse gases. The energy generated by an ESBWR will prevent the emission of approximately 7.5 million metric tons of CO2 per year compared to standard electricity production on the U.S. grid. The analysis present in this paper aimed to characterize the thermal-hydraulic simulations of full-scale ESBWR design. The analysis presented will help in recognizing the improvement needed in the reactor design and its passive safety systems. The analysis is performed for normal steady-state and postulated design basis accident scenarios (break in one of the reactor main-steam lines inside the contentment). The simulation results obtained by the code REALP/SCDAPSIM/MOD3.4 are compared with the TRACG04 and MELCOR2.1 code results to determine the code predictability and accuracy under accident conditions of the newly proposed design of the ESBWR nuclear reactor. A parametric study is also carried out with different in-containment break sizes to determine code results' sensitive nature due to natural circulation flow phenomena, which are highly unstable. The analysis shows that the code has produced reasonably accurate results under the tested system thermal-hydraulic conditions compared with the TRACG04 and MELCOR2.1 code simulation results. This agreement between the three codes simulation results demonstrated that the publicly available RELAP5 code model has adequate features to simulate the ESBWR thermal-hydraulics. It has been also demonstrated that for the postulated accident conditions, the design of passive safety systems is capable to capture the accident progression without any active power.

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

A probabilistic fracture mechanics (PFM) analysis code PASCAL was developed in Japan for probabilistic integrity assessment of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. To strengthen the practical applications of PFM methodology in Japan, PASCAL has been upgraded to a new version, PASCAL5, which enables PFM analyses of RPVs in both PWRs and boiling water reactors (BWRs) subjected to a broad range of transients, including PTS and normal operational transients. In this paper, the recent improvements in PASCAL5 are described such as the incorporated stress intensity factor solutions and corresponding calculation methods for external surface cracks and embedded cracks near the RPV outer surface. In addition, the analysis conditions and evaluation models recommended for PFM analyses of Japanese RPVs in BWRs are investigated. Finally, PFM analysis examples for core region of a Japanese BWR-type model RPV subjected to two transients (i.e., low-temperature over pressure and heat-up transients) are presented using PASCAL5.

A numerical investigation into Metallic-Melt continuous drainage in the core support plate region of a BWR for the initial phase of core melt progression

Under a typical uncovered core scenario in a Boiling Water Reactor (BWR), it is not clear that a large core molten pool will result as occurred in the Three Mile Island, Unit 2 (TMI-2) accident. Sandia National Laboratories (SNL) has propounded the alternative “Continuous Drainage” melt progression for the scenario, where molten material drains from the core region without the formation of any stable crusts or blockages. Hence the present simulation study was conducted to investigate the behavior of the metallic-melt continuous drainage and the drainage pathways during the initial phase of core melt progression in a BWR under an uncovered core condition like that encountered in the Fukushima Daiichi Nuclear Power Plants (NPPs) accident experienced by Tokyo Electric Power Company Holdings (TEPCO). Researchers around the world have conducted numerous sophisticated and comprehensive studies using simulation codes and experiments to investigate the Fukushima accident. Melt drainage with deforming components is subject to many uncertainties in a core support plate region possessing rather complicated geometries. Additionally, there is very limited experimental and numerical simulation knowledge on melt drainage in such a region. Thus, in recent years, Japan Atomic Energy Agency (JAEA) has experimented on melt drainage using a simplified apparatus to reproduce the behavior. The present paper provides simulation results on melt drainage in the core support region using SAGITTARIUS, an in-house three-dimensional simulation code capable of consistently simulating melt drainage with crust formation, local blockage, and structural damage. The in-vessel simulation target with a prototypic BWR-composition and -geometry is composed of four fuel assemblies (4 × 4 fuel rods per assembly), an absorber blade, a support piece, a core support plate, an Instrument Guide Tube (IGT), and a Control Rod Guide Tube (CRGT) with a Control Rod Drive Housing (CRDH). We estimated the drainage speed for the JAEA experiment to validate the SAGITTARIUS function. The drainage speed was evaluated to be 3.1 m/s, a speed corresponding to that observed in the core support plate region. We also used SAGITTARIUS to simulate the behavior of the metallic-melt drainage and temperature rise of the structure. The results from SAGITTARIUS on the trend and values of the temperature rise agreed with the measured values. On the basis of this finding, we conducted numerical investigations on the metallic-melt drainage of B4C-Stainless Steel (SS) eutectic from the absorber blade and Zry-melt from the fuel rods. The simulation demonstrated the capacity to describe the following features: (i) visual demonstration of metallic-melt drainage from the absorber blade and fuel rods, (ii) aggressive Zry-melt attack to the inlet orifice, (iii) crust formation with local blockage in the narrow drainage paths, and (iv) sedimentation of solidified melt on the bottom faces. These findings were consistent with the concept of the “Continuous Drainage” proposed at SNL and we supported the SNL scenario by the numerical investigation. These investigations probably provide useful knowledge during the discharge of damaged fuels or crust in the Fukushima-Reactor Pressure Vessels (RPVs).

Adverse Effects of Condenser Cooling Seawater Temperature, Fouling, and Salinity on the Output Power and Thermal Efficiency of BWR NNPs

Increasing the thermal efficiency in newly designed power stations is a priority. Keeping the efficiency in existed plants close to the rated one is of paramount importance. This research contributes to investigating the adverse effects of changes in condenser seawater coolant characteristics, (temperature, fouling, and salinity), on the thermal performance of a Boiling Water Reactor Nuclear Power Plant (BWR) NPP. A mathematical model is developed to relate seawater cooling temperature, fouling, and salinity to output power and thermal efficiency. The model also explains the impact of the condenser performance on power and efficiency. The thermal efficiency of the considered BWR NPP is reduced by 2.26% for a combined extreme increases in the condenser cooling seawater temperature, fouling factor of seawater and treated boiler feed water, and salinity by 10 °C, 0.0002, 0.00001 m2K/W, and 100 g/kg, respectively. A rise in the condenser efficiency from 40 - 100 % results in an increase in the output power by 7.049%, and the thermal efficiency increases by about 2.62%. Conclusions are useful for reactor’s design

Experimental investigation of solid particle spreading driven by gas injection into a pool of water

Motivated from the fuel–coolant interaction phenomena in boiling water reactors, in the present work, effect of natural convection flows, set during the melt dripping, on the nature of formation of debris bed, of solidified particles, at the bottom floor is studied. Standard shape solid particles are used to simulate the dripping melt and their paths are tracked using a particle tracking technique to acquire additional data such as particles velocity, travel time and path. The experimental studies performed on PDS-P facility are designed to study the separate effects and generate the data for codes validation. A novel particle tracking technique allowed quantification of kinetic properties for every particle. The results helped in refinement of the particles distribution on the floor, quantifying the debris bed shape. Higher pool depths and natural convection flows rates are seen effective in enhancing the distribution of debris particles, creating shallow and well spread debris bed.

Evaluation of the Structural Integrity of a Boiling Water Reactor Core Shroud with an Irregular Distribution of Circumferential Cracks: Elasto-plastic and Plastic Analysis

Three asymmetric cracks were postulated at the circumferential horizontal weld H7 of a boiling water reactor type 5 core shroud after 25 years of operation. They were arranged in an irregular distribution with different depths. Unstable crack propagation can be induced when an earthquake and a loss of coolant accident take place. In this paper, a ductile failure analysis has been reported. It has been done following an elastoplastic analysis and a plastic limit load analysis. The crack conditions after 30 and 60 years of operation were evaluated. The results show that the core shroud will keep its structural integrity during sixty years under normal conditions of operation. On the other hand, its operational life is reduced five years under failure conditions. In general terms, the methodology, which has been developed, can be applied for any number of circumferential cracks located in the same cross section. Fast and reliable outputs can be obtained. It can be applied in any moment of the life of the plant, to support the decision on “Use as is”, “Mandatory Repair”, or “Estimation of the remaining life.”

Application of uncertainty analysis methods to MELCOR simulation of postulated severe accidents in a Nordic BWR

Different uncertainty analysis methods are applied to MELCOR simulation of two postulated severe accidents in a Nordic boiling water reactor (BWR): (i) station blackout (SBO) accident, and (ii) large break loss-of-coolant accident (LBLOCA) combined with SBO, with the objective to compare their performances in the estimates of 95/95 tolerance limits of two figures of merit (FOMs) – the hydrogen mass produced from core degradation and the timing of vessel failure. Given 17 uncertain input parameters of MELCOR with probability density functions (PDFs), the 95/95 estimates of the two FOMs are obtained through the uncertainty analysis. From the uncertainty analysis results, it is found that for the quantification of single FOM a larger sample size leads to a much more accurate and stable 95/95 estimate at a higher computational cost, and the three nonparametric methods (Wilks’ method, Beran and Hall’s linear interpolation method as well as Hutson fractional statistics method) behave similarly in both accidents, while the goodness-of-fit test method performs differently and tends to provide a more realistic 95/95 estimate in both accidents. For the quantification of multiple FOMs the bracketing method tends to provide a smaller 95/95 estimate than the Wald and Guba method does, in consistent with their mathematical definitions. The Wald and Guba method is more stringent than the bracketing method when all percentiles (coverage) are set as the same. The sensitivity analysis results show that the several most significant input parameters are ranked almost identically by Spearman rank correlation coefficient (SRCC) and Pearson correlation coefficient (PCC), but these coefficients are dependent on accident scenarios and output parameters. Among the 17 parameters chosen, molten cladding drainage rate is the most influential to the output parameters (timing of initial melt relocation, timing of vessel failure, residual heat, etc.) considered in the present study, probably due to its impacts on molten Zr exposure to steam and resulting oxidation.

Formulation of Irradiation Assisted Stress Corrosion Crack Growth Rates for Neutron-Irradiated Stainless Steels under Hydrogen Water Chemistry Conditions of Boiling Water Reactors

ABSTRACT This study developed a formula for the calculation of crack growth rates (CGRs) of irradiated austenitic stainless steels (SSs) under hydrogen water chemistry (HWC) conditions of boiling water reactors. First, the input parameters for the Hashimoto–Koshiishi model, which is a mechanistic crack growth model of stress corrosion cracking (SCC), were optimised for SSs irradiated up to 11 dpa. Electrochemical corrosion potential (ECP) of those SSs during crack growth tests nearly equalled −200 mVSHE. Second, a simplified formulation was developed to reproduce the model calculations of SSs. Calculated CGRs agreed well with experimental CGRs of various types of SSs, which were irradiated up to 13 dpa and tested at ECPs from −550 mVSHE to −200 mVSHE, when the experimental CGRs were less than the values for NWC conditions. On the other hand, there were some data in which experimental CGRs were much higher than calculated CGRs. By analysis of the experimental conditions of the data, the possible influencing factor for the discrepancy between the calculated and experimental CGRs was found to be the -validity. If the -validity criterion was not satisfied, the strain distribution used in the model calculation was not adequate for the actual strain distribution.

Analysis of Fukushima Daiichi unit 4 spent fuel pool using MELCOR

A spent fuel pool is a temporary storage facility designed to store the nuclear fuel assemblies removed from the reactor core. Here, the depleted fuel is vertically arranged in racks structures, accurately spaced to ensure subcriticality safety margins. Until the events occurred at the Fukushima Daiichi power plants, severe accidents in spent fuel pools had never been considered as a major safety concern, mainly because of their slow progression. However, extremely unlikely accident events, aggravated by the loss of site power, highlighted the vulnerability of nuclear fuel stored within spent fuel pools. During these accident sequences, reactor fuel rods can generate enough decay heat to vaporize the coolant around them, causing, during unmitigated scenarios, the uncovering and melting of fuel bundles after several days. In past years, new features have been introduced in the MELCOR code for the evaluation of potential risks associated to accident progression in spent fuel pools. In this paper, the accidental sequence of a loss of cooling accident in a boiling water reactor spent fuel pool is analysed with version 2.2 of the MELCOR code. The spent fuel pool of the Fukushima Daiichi unit 4 has been selected as reference. Main accident events, such as fuel uncovering, claddings oxidation, hydrogen generation, and fission products release, have been analysed assuming the unavailability of safety systems.

Aircraft-structure interaction study for the assessment of the reaction-time response of nuclear containment structure

A systematic finite element (FE) investigation has been presented in this study to basically examine the two basic constraints (deformation and curvature) in the conventional reaction-time response approach for studying the behavior of a nuclear containment structure. A coupled finite element investigation has been carried out wherein the reaction-time response of the Boeing 707–320 and Boeing 747–400 has been obtained against the rigid flat and curved targets of diameter 42 and 100 m in order to study the possible influence of curvature on the reaction-time response curve. The aircrafts were subsequently considered to hit the deformable containment of Boiling Water Reactor Nuclear Power Plant (BWR-NPP) of diameter 42 m. The effect of target deformability thus obtained in the reaction-time response has been compared with that of the one obtained against the rigid target of the same curvature. Further, the numerical as well as analytical approaches have been employed to evaluate the exact impact area at the interface of the aircraft and flat rigid target as a function of time and incidence velocity. Finally, the behaviour of BWR-NPP has been studied against Boeing 707–320 and Boeing 747–400 using the coupled approach, traditional average contact area approach and area trifurcation (fuselage, first set engines and second set engines) approach. The response of the containment thus obtained against the three approaches has been compared.

Impact of Chloride on the Environmentally-Assisted Crack Initiation Behaviour of Low-Alloy Steel under Boiling Water Reactor Conditions

Low-alloy reactor pressure vessel steels have a rather low susceptibility to stress corrosion cracking (SCC) in a boiling water reactor (BWR) environment if the high-temperature water contains no anionic impurities. Recent investigations revealed that under oxidizing BWR normal water chemistry (NWC) conditions extremely small amounts of chloride, can cause very high SCC growth rates in these materials. Therefore, the effect of continuous and temporary chloride additions on the crack initiation behaviour was explored by a series of constant extension rate tensile (CERT) and constant load tests in high-temperature water. In an NWC environment, containing ≥2 ppb of chloride, strain-induced corrosion cracking (SICC) initiation occurred briefly after the onset of plastic yielding and at much smaller strains than in high-purity water. On the other hand, under reducing hydrogen water chemistry conditions with up to 700 ppb chloride, no SICC was detected up to very high strains. CERT experiments, with moderate short-term chloride transients before and during the loading, showed that even serious mechanical loading transients, one day after returning to high-purity water, did not result in early SICC initiation.

Generic and adaptive probabilistic safety assessment models: Precursor analysis and multi-purpose utilization

Motivated by learning from experience and exploiting existing knowledge in civil nuclear operations, we have developed in-house generic Probabilistic Safety Assessment (PSA) models for pressurized and boiling water reactors. The models are computationally light, handy, transparent, user-friendly, and easily adaptable to account for major plant-specific differences. They cover the common internal initiating events, frontline and support systems reliability and dependencies, human-factors, common-cause failures, and account for new factors typically overlooked in many PSAs. For quantification, the models use generic US reliability data, precursor analysis reports, the ETHZ Curated Nuclear Events Database, and experts’ opinions. Moreover, uncertainties in the most influential basic events are addressed. The generated results show good agreement with assessments available in the literature with detailed PSAs. We envision the models as an unbiased framework to measure nuclear operational risk with the same “ruler”, and hence support inter-plant risk comparisons that are usually not possible due to differences in plant-specific PSA assumptions and scopes. The models can be used for initial risk screening, order-of-magnitude precursor analysis, and other research/pedagogic applications especially when no plant-specific PSAs are available. Finally, we are using the generic models for large-scale precursor analysis that will generate big picture trends, lessons, and insights.

Mitigation of long-term station blackout accident of advanced boiling water reactors

Taiwan Power Company proposed ultimate response guidelines (URGs) to mitigate the so-called ‘Fukushima type accident’ of nuclear power plants. As stipulated in URGs, operators will depressurize the reactor coolant system if necessary to bring in low-pressure water. Containment venting is then used to dissipate the heat generated from the decay. These collective actions are called DIVing (depressurizing, injecting and venting) in URGs. The actions are designed to avoid large-scale evacuation of the public around the damaged plant during an accident. In this study, the effectiveness of DIVing in preventing radioactive emission to the environment is evaluated using RELAP5 codes. The plant analyzed is Lungmen Nuclear Power Station, which has two 1350-MWe advanced boiling water reactors (ABWR) designed by General Electric. In the Standard Operating Procedure of DIVing in URGs, controlled depressurization is executed to bring the pressure to the lowest operating pressure of Reactor Core Isolation Cooling (RCIC) system. The core water level is kept high while RCIC is in operation.. Emergency depressurization is carried out when RCIC fails. The executing time of DIVing as specified in URGs is too conservative. The study identifies the acceptable initial states for operators to carry out DIVing actions. The success criterion is to keep the peak cladding temperature below 1088 K (1500°F) during the transient, which is the temperature of cladding breach and the initiation of gap release of radionuclide from fuel rod.In this study, depressurization transients in the station blackout accident of ABWR are simulated using RELAP5/3D. The results show that the peak cladding temperature obtained during a depressurization transient depends on the time of depressurization, initial pressure, initial water level and injection rate of low-pressure water. When the water level and injection rate are fixed, a region of successful depressurization on the depressurization time vs pressure plot is delineated based on the results of RELAP5 simulations. Regions of successfully depressurization for typical combinations of injection rate and water level are identified, and acceptable region’s boundary is linear.A computation aid, Guidance for Emergency Depressurization (GfED), was developed based on the simulation results of the RELAP5/3D code to assist operators in determining whether the emergency depressurization under given conditions can be performed successfully. GfED uses low-pressure injection flow rate, water level and initial pressure upon emergency depressurization as input parameters to calculate the latest time that is acceptable for emergency depressurization. The technique of linear interpolation is adopted. By comparing the result and value input by the operators, GfED can predict whether it is safe to depressurize.GfED also guides operators to determine the initial condition of a successful depressurization. Three of the four input parameters are fixed in the guidance, and the limiting value of the fourth parameter of a successful depressurization is calculated. These values are informative for operators to manoeuvre the system while the reactor core isolation cooling system is still running.

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Abstract The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory, etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, and choice of calculation parameters. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of serpent. The effect of the following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy-dependent branching ratios. As a test case, a 2D Boiling Water Reactor (BWR) fuel assembly was modeled by first running a burnup calculation followed by a decay calculation. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general, the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6% and for the photon emission rate approximately 1.1%. For the spontaneous fission rate maximum absolute relative difference of nearly 8% was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy-dependent branching ratios had an insignificant effect on the studied source term components.

Experimental Investigation on the Retention of Soluble Particles by Pool Scrubbing

Abstract In the late stage of a severe loss-of-coolant accident, the pressure in the containment building of a nuclear power plant could rise beyond the design limits and thus endanger its structural integrity. Therefore, to avoid pressure failure, it may be necessary to perform controlled venting of the containment. During the event of an accident, a large amount of fission and activation products are released into the containment as airborne particles (aerosols). These particles are filtered during the venting process, usually with the help of wet filters, in order to keep risks to the surrounding environment to a minimum. Consequently, the knowledge of the retention processes in a water reservoir (pool scrubbing) is of central importance for such filtered containment venting systems and for reactor concepts in which water reservoirs are used for pressure reduction (e.g., condensation chamber of a boiling water reactor (BWR)). Investigations on pool scrubbing are carried out in the severe accident aerosol behavior (SAAB) test facility at the Juelich Research Center in the framework of a national project (“severe accident aerosol behavior-II”). SAAB is a unique large-scale facility with the ability to perform a great variation of experiments using various measurement tools. The influence of numerous parameters, such as the height of the water pool, solubility of aerosols and concentration on the retention capacity, is investigated by means of separate effect studies on both insoluble and soluble particles. This technical brief gives a detailed overview over the facility and includes example results of the first test series with soluble particles including cesium iodine (CsI).

Appraisal of the NRC H-uptake calculation for Swiss boiling water reactors

Uptake of hydrogen by the cladding of fuel rods during LWR operation worsens the mechanical properties of the material and can lead to cladding failure. The U.S. NRC has introduced a dependency of the equivalent cladding reacted (ECR) criterion on the H-content in the cladding and has proposed a burnup dependent calculation model of the H-content based on a large database which does not cover high burnup such as those reached in Switzerland. This work addresses how the extrapolation of the proposed H-content calculation is conservative compared to the experimental values on claddings irradiated in Swiss plants. Therefore, a model is proposed to calculate the H-content in Swiss BWR fuels up to high burnup, based on published experimental data. The new correlation leads to less conservative results. This is demonstrated with the simulation of a postulated LOCA in a Swiss BWR, where the high exposure bin in the core is no longer expected to fail.

Phenomenological model of disturbance waves in annular two-phase flow

A new, phenomenological, model of disturbance waves in annular two-phase flow is presented. The model builds upon the three-field (gas, drops and film) one-dimensional model of annular two-phase flow by considering the disturbance waves as an additional fluid field, traveling over a thin liquid base film, along with its own mass, momentum and energy conservation equations. The waves are modeled as fluid particles characterized by basic macroscopic parameters such as the wave shape factor (amplitude/width) and the wave number density (spatial frequency of waves). Phase change is considered and relevant closure models are developed for the field mass and momentum exchange rates (drop entrainment/deposition and waves/base film interactions). A Boltzmann transport equation of wave number density is introduced to model the hydrodynamic non-equilibrium effects related to the wave creation, merging and splitting processes, characteristic of developing annular two-phase flows (due to e.g., inlet effect, phase change, geometrical change or transient). Finally, the base film mass and wave number density sink and source rates are modeled using a relaxation time approximation. The new model is motivated and fully documented in this paper and has been implemented in the latest version of the transient sub-channel analysis code MEFISTO-T. While the proposed model framework is generic for coherent disturbance waves, the required wave characteristic correlations are specifically developed and calibrated using high pressure steam/water experiments under equilibrium conditions relevant of Boiling Water Reactor operation. The model relaxation times are then estimated by comparing the model predictions against measured wave parameters downstream various flow perturbations (evaporation and film suction), along the channel recovery length toward equilibrium. For all considered experiments, the model predicts the averaged wave and base film parameters typically within +/-20% and is able to capture the axial development of disturbance wave characteristics. The consideration of non-equilibrium effects is found fundamental to the accurate simulation of axially developing annular two-phase flows.

Variational mode decomposition method (VMD) applied to Decay Ratio (DR) calculation for instabilities identification in BWR

Decay Ratio (DR) is the stability indicator most used in Boiling Water Reactors. Its calculation can be performed by several methods which can lead to errors due to the fact that instability time series are not stationary. In the last years, several authors have proposed to use decomposition methods such Empirical Mode Decomposition (EMD) or Ensemble Empirical Mode Decomposition (EEMD) to obtain mono-frequency modes from the original signal. These techniques, although useful, produce mode mixing and they are not completely valid for an on-line implementation. Variational Mode Decomposition (VMD) is a recently developed algorithm which solves the mode mixing problem. In this work it has been used to improve the estimation of DR: Comparisons with traditional methods are also included.

Preliminary evaluation of decommission waste inventory and quantity for nuclear power plants

<p>Due to the Taiwan national policy ""nuclear-free homeland by 2025"". As a result, Taiwan nation’s three operating Nuclear Power Plants (NPPs) are being decommissioned. </p> <p> Decommissioning is a representative stage in back-end cycle of NPP. There has been several NPP decommissioning indicated Global Warming Potential (GWP) could also be reduced by reducing the volume of waste and increasing recycling of material and steel. The Taiwan nation’s three operating nuclear NPPs are to be decommissioned and has progressed the strategy determination and Research and Design (R&D) for the decommissioning. </p> <p> Decommissioning waste, components, system and structure is one of the most important elements. Decommissioning waste quantity is calculated based on Piping and Instrumentation Diagram Piping (P&ID) and completion report with activation and contamination data. Due to the largest portion of waste management and disposal in decommissioning, it is necessary to exactly evaluate waste quantity (applying the regulation, guideline, and site-specific characterization) for economic viability.</p> <p> The purposes of this paper are to provide an overview of useful as an important information in establishing the waste inventory and quantity for Boiling Water Reactors (BWR) are mainly described.</p> <p> </p>

Validation of the TWOPORFLOW code for the core analysis of liquid metal-cooled reactor with selected experiments

TWOPORFLOW (TPF) is a thermal–hydraulic code that utilizes a porous medium approach in three-dimensional Cartesian coordinates. It can simulate the flow phenomenon in fuel assemblies or reactor cores by solving three conservation equations (single-phase) or six conservation equations in a two-fluid liquid–vapor approach. Many constitutive correlations e.g. pressure loss, interfacial friction, and heat transfer (wall, liquid–vapor interface), are included to close the system of equations. The code was recently extended to simulate pre-Critical Heat Flux (CHF) and CHF heat transfer relevant for Boiling Water Reactor (BWR). In this paper, the code's simulation capability for liquid metal-cooled reactors core is validated using three experimental data sets: pressure drop of the 19-pin rod in THEADES LBE loop, the BREST-type reactor benchmark problem, and the NACIE-UP Block Fuel Pin Simulator (BFPS) test. The results show that TPF can be used not only for square rod bundle arrangements but also for hexagonal arrangements; the code can reasonably predict the thermal–hydraulic behavior of liquid metal-cooled reactors under normal and off-normal conditions.