Boiling Water Reactor Nuclear Power Research Articles

The structural and non–linear dynamic analysis for radioactive waste container

In recent years, the development of radioactive waste containers for nuclear facility decommissioning and dismantling is a critical issue because the Taiwan domestic boiling water reactor nuclear power plant is going to be decommissioned. The main purpose of this research is to design a metal container that meets the structural requirements of related regulations. At first, the shielding analysis was performed by varying dimensions of radioactive waste to determine the storage efficiency of the container. Then, a series of structural analyses for operational and accidental conditions of the container with full load were conducted, such as lifting, stacking, and drop impact conditions. On the other hand, the field drop impact tests were carried out to ensure structural integrity. The present research demonstrates the structural safety of the developed container for decommissioned nuclear facilities in Taiwan.

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

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.

A Newly Developed Suppression Pool Model Based on the ISAA Code

The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas flows from the drywell to the wetwell after the normally submerged connecting vents between the drywell and the wetwell have been purged of water through a vent clearing process so that the suppression pool may prevent pressure in the containment from exceeding the designed pressure limit. To analyze such complicated thermal-hydraulic behavior of the suppression pool under a specific accident, an advanced reasonable model should be developed. The SPARC-90 model used in MELCOR calculates the mass and energy transfer between the bubbles of the injected gas and the suppression pool, which is affected by distance efficiency and subcooling efficiency. The dedicated vent flow model used in CONTAIN can well simulate the vent clearing time, that is, the time required for the liquid level on either side of the suppression pool to drop to the level at which the vent begins to clear. However, it is necessary and better to combine them into one integrated code. This paper presents a newly developed suppression pool model based on the self-developed severe accident analysis code Integrated Severe Accident Analysis (ISAA), which combines the advantages of the dedicated vent flow model and the SPARC-90 model to analyze the suppression pool’s thermal-hydraulic behavior. The simulation results of the developed suppression pool model shows reasonableness compared with the result in the CONTAIN 2.0 code manual. The good agreement between the simulation results and the analysis results from the COLUMBIA power station final safety analysis report demonstrates the rationality and effectiveness of the developed model, although future improvement is needed.

Effect of phosphorus content on stress corrosion cracking susceptibility of shielded metal arc weld metals for 600 type alloy in high temperature pressurised pure water*

The intergranular stress corrosion cracking (IGSCC) susceptibility of shielded metal arc weld metals for Inconel alloy 600 has been investigated with particular reference to the influences of P and Nb contents on the basis of a creviced bent beam (CBB) test in pressurized hot water (corresponding to the service condition of boiling water reactor nuclear power plant). The IGSCC susceptibility of the weld metals in the as-welded state exhibited a tendency to increase with P content. When the post-weld heat treatment stress relief+low-temperature ageing (SR+LTA) of SR annealing at 893 K followed by ageing at 673 K was applied, however, the influence of P content on the IGSCC susceptibility of weld metals depended on the Nb content; i.e. the IGSCC susceptibility of the weld metals containing 1.2 mass% Nb was almost independent of P content, whereas the weld metal containing 4.4 mass% Nb showed much decreased IGSCC susceptibility by increasing the P content from 0.012 to 0.031 mass%. Since the double-loop electrochemical potentiokinetic reactivation test suggested that the grain boundary corrosion resistance of the weld metals of 4.4 mass% Nb decreased with the rise in P content, the improved IGSCC susceptibility of the weld metals of 4.4 mass% Nb and 0.031 mass% P cannot be explained in terms of the corrosion resistance of the grain boundary. The weld metal containing 4.4 mass% Nb was hardened significantly by the precipitation of the γ′ phase (Ni3Nb) in interdendritic areas during the SR+LTA treatment. The increase in the hardness brought about by the SR+LTA treatment was depressed significantly by the rise in P content probably due to the reduction in the precipitation area of the γ′ phase. It can be considered that the decreased hardness contributes to the improvement in the IGSCC susceptibility with increasing the P content, since the tensile stress applied to the specimen during the CBB test increases with the hardness.

Effect of chromium content on stress corrosion cracking susceptibility of shielded metal arc weld metals for 600 type alloy in high-temperature pressurised pure water

The stress corrosion cracking (SCC) susceptibility of SMAW metals for Inconel alloy 600 to which Cr was added to 14.8–21.4 mass% has been investigated on the basis of a creviced bent beam test in pressurized hot water (corresponding to the service condition of boiling water reactor nuclear power plant), since the TIG weld metal of alloy 82 involving 18–22 mass% Cr possesses much better resistance to SCC than the SMAW metal of alloy 182 (Cr content = 3–17 mass%). When their Cr contents were increased to the same level as those of the alloy 82, the weld metals of alloy 182 sustained only slight SCCs in the as-welded state, and no crack was detected after the post-weld heat treatment stress relief+low-temperature ageing (SR+LTA) of stress relief annealing at 893 K followed by ageing at 673 K. These results suggest that the higher Cr content of the alloy 82 is responsible for its higher resistance to SCC than that of the alloy 182. The Cr carbides precipitated at the grain boundary during the welding, and the SR+LTA treatment was also changed from M7C3 type to M23C6 type with the increase in the Cr content. Although the Cr content at the grain boundary in weld metal containing 14.8 mass% Cr subjected to the SR+LTA treatment was 3 mass%, the Cr content of weld metal containing 18.5 mass% Cr was not less than 10 mass%. The addition of the Cr to the alloy 182 increased the Cr content in the grain boundary region, suggesting that the inter-granular SCC can be suppressed when the Cr content at the grain boundary is not less than 10 mass%. In addition to the carbide, Ni16(Mn, Cr)6Si7 (G phase) was precipitated at the grain boundary in the alloy 182 containing 18.5 mass% Cr when the SR+LTA treatment was applied. transmission electron microscope-EDS analyses suggested that the G phase was enriched in P, and therefore could decrease the P content in the grain boundary region. Probably, the decreased P content at the grain boundary due to the precipitation of G phase contributed to the enhancement of the SCC resistance of the Cr-added alloy 182 by the SR+LTA treatment.

The Pure Heat Conversion Coefficient Analysis Method for Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit

Pure heat conversion coefficient, which is one important parameter for the advance boiling water reactor nuclear power unit, is defined, and the coefficient reflects energy grade values of heaters. According to the structural characteristics of the thermodynamic system of the advanced boiling water reactor nuclear power plant, four sorts of auxiliary steam-water components are categorized. Via strict deduction and demonstration, the general matrix of the coefficient is deduced, so the thermal economic analysis of advance boiling water reactor nuclear power unit can be done with one of extraction steam efficiency. In this way, a new method of thermal economic quantitative analysis for this unit is offered.

Decay Ratio Estimation of BWR Signals Based on Wavelet Ridges

A wavelet ridge application is proposed as a simple method to determine the evolution of the linear stability parameters of a boiling water reactor nuclear power plant (NPP) using neutronic noise signals. The wavelet ridges are used to track the instantaneous frequencies contained in a signal and to estimate the decay ratio (DR). The first step of the method consists of denoising the analyzed signals by a discrete wavelet transform to reduce the interference of high-frequency noise and concentrate the analysis in the band where crucial frequencies are presented. Next is computation of the wavelet ridges by a continuous wavelet transform to obtain the modulus maxima from the normalized scalogram of the signal. In general, associations with these wavelet ridges can be used to compute the instantaneous frequency contained in the signal and the DR evolution with the measurement. To study the performance of the wavelet ridge method, by computing the evolution of the linear stability parameters, both simulated and real neutronic signals were considered. The simulated signal is used to validate methodically and to study some features of the wavelet ridge method. To demonstrate the method applicability, three real neutronic signals related to instability events in the Laguna Verde NPP and Ringhals and Forsmark stability benchmarks were analyzed. The investigations show that most of the local energies of the signal are concentrated and that DR variations of the signals were observed along the measurements.

Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes

A very complex type of power instability occurring in boiling water reactor (BWR) consists of out-of-phase regional oscillations, in which normally subcritical neutronic modes are excited by thermal-hydraulic feedback mechanisms. The out-of-phase mode of oscillation is a very challenging type of instability and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, simulations of out-of-phase instabilities in a BWR obtained by assuming an hypothetical continuous control rod bank withdrawal are being presented. The RELAP5/Mod3.3 thermal-hydraulic system code coupled with the PARCS/2.4 3D neutron kinetic code has been used to simulate the instability phenomenon. Data from a real BWR nuclear power plant (NPP) have been used as reference conditions and reactor parameters. Simulated neutronic power signals from local power range monitors (LPRM) have been used to detect and study the local power oscillations. The decay ratio (DR) and the natural frequency (NF) of the power oscillations (typical parameters used to evaluate the instabilities) have been used in the analysis. The results are discussed also making use of two-dimensional plots depicting relative core power distribution during the transient, in order to clearly illustrate the out-of-phase behavior.

Effectiveness of the Raw-Water System as a Severe-Accident-Management Measure in Station Blackout Accidents

Chinshan is a Mark-I boiling water reactor nuclear power plant (NPP) located in north Taiwan. It incorporates several severe-accident-mitigating features, especially two raw-water tanks in the mountain. According to a probabilistic risk assessment (PRA) of Chinshan NPP, station blackout (SBO) sequences are the most dominant sequences in internal core damage frequency. No credit is taken for the raw-water system in the development of a Chinshan PRA. Therefore, two dominant sequences (T3UTERDGX and T3UTERDG) of the SBO in the Chinshan PRA are cited as reference cases to evaluate the capacity of the raw-water system in the PRA and severe accident. The T3UTERDGX sequence is initiated by loss of off-site power (T3) followed by failure of both diesel generators (DG), failure of gas turbine generators, and failure to recover alternating current (ac) power (ER). That results in loss of all on- and off-site ac power. The high-pressure injection systems fail (UT) initially and timely reactor depressurization fails (X). The T3UTERDG sequence is the same as the T3UTERDGX sequence, except for failure of timely reactor depressurization (X). The MAAP4 code is used as a tool to evaluate the effectiveness of the raw-water system. Based on MAAP4 analysis, the raw-water system cannot cool down the core in the T3UTERDG sequence after introducing severe-accident-management guidelines. The raw-water system cannot flood dry-well water level above minimum debris submerge level (MDSL) in the T3UTERDGX sequence after reactor pressure vessel (RPV) breach. Sensitivity studies show that raw-water injection before the vessel water level reaches level 2 (L-2) can keep core coolability in the T3UTERDG sequence. Three times the raw-water injection rate is the minimum flow rate to flood the dry-well water level above MDSL and cool down the corium on the dry-well floor in the T3UTERDGX sequence. A raw-water system can be used as a mitigating measure, especially in an SBO. The RPV should be depressurized as quickly as possible if a raw-water system is the only mitigation measure in the accident. It is worthwhile to increase the raw-water flow rate to cool down the debris in the dry well after RPV breach.

Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor’s power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

The study of CRUD removal resin XT-1033

In order to reduce the occupational radiation load at BWR (Boiling Water Reactor) and ATR (Advanced Thermal Reactor) type Nuclear Power Plants, it is important to improve the iron removal efficiency at the CD (condensate demineralizer) of feed water system.The CRUD removal efficiency of CD depends on the cation exchange resin's ability which is known to be improved with time.The newly developed cation exchange resin “XT-1033” shows a excellent CRUD removal efficiency in the study of applocation at actual plant from fresh age.

Advanced Control Complex for BWR Nuclear Power Plant

A third generation control and operation system A-PODIA has been developed for Boiling Water reactor nuclear power plants. It is a highly automated monitoring and operation system which has a compact and elegant human interface. The functions and performance of A-PODIA and the future control systems are reviewed.

Application of RETRAN-02 to Plant Transient Analysis and Issue Resolution

The applicability of a complete and well-qualified plant transient code and model to support in-depth evaluations of anomalous plant transients is shown. Analyses of several operational and abnormal transients that occurred during the first three cycles of the Cofrentes boiling water reactor nuclear power plant are presented. Several safety issues for the Cofrentes and Almaraz power plants are also resolved. These applications demonstrate the need for a very detailed and accurate simulation of the control systems as well as the convenience of having a data acquisition system that is as complete as possible.

Experience with the Use of Recycle Plutonium in Mixed-Oxide Fuel in Light Water Reactors in the Federal Republic of Germany

In the Federal Republic of Germany (FRG), the reprocessing of spent fuel assemblies and the recycling of the recovered plutonium are essential for the final disposal of high-level nuclear waste. In addition, the recycling of uranium and plutonium is an effective measure to avoid proliferation of plutonium for weapons and to reduce the needs of uranium. Therefore, the technology of recycling has been developed in the FRG since 1966. In the prototype boiling water reactor (BWR) of Kahl, in the BWR nuclear power plant at Gundremmingen, and the pressurized water reactor plant at Obrigheim mixed-oxide fuel (MOX) assemblies have been operated with good success. Also, the experience gained at Alkem's MOX fuel fabrication plant is very positive. Reprocessing tests of spent MOX fuel rods have indicated that a nearly homogeneous mixture of PuO/sub 2/ and UO/sub 2/ should be used to achieve a good solubility of MOX fuel in nitric acid. In all, very important experience in safeguarding plutonium and MOX fuel material was gained. 7 figures, 3 tables.

Special external radiation field study at the Vermont Yankee Nuclear Power Station.

A brief field study was undertaken at the Vermont Yankee Nuclear Power Station during January-April 1973 at the request of the Vermont Division of Industrial Hygiene. The purpose of this study was to demonstrate the use of thermoluminescent dosimeters, a pressurized ionization chamber, and portable survey instruments to measure external gamma radiation on which to base population exposure estimates in the vicinity of a boiling water reactor nuclear power station. The station, located at the southeastern corner of Vermont, is a 540 MW(e) BWR. Results of this study include: (1) pressurized ionization chamber measurements at locations of critical population exposure to detect small increments of radiation esposure (0.08 mR in 18 days) due to the plume of radioactive gases; (2) measurements of direct radiation from the station at levels of 0.4 to 1 μR/hr using portable instruments; (3) thermoluminescent dosimeter measurements for periods of 2–3 weeks to detect environmental radiation exposures due to station operations: and (4) identification of long-term variations in background due to snow and/or location for evaluating environmental surveillance data at or near background levels. The observations demonstrate the availability of techniques for measuring radiation exposures of several mR/yr in the vicinity of a boiling water reactor. A program at these low levels requires careful se1ection of monitoring locations, detailed examination of the natural radiation background, and thorough quality assurance for the measurements.

131-I discharges from an operating boiling water reactor nuclear power station.

A series of radiological surveillance studies was conducted at Dresden Nuclear Power Station by the Radiological Engineering Laboratory of the Division of Environmental Radiation which provided data on 131I concentrations observed in the plant, the plant's effluent, and the environment. Based on the Dresden study data, an evaluation is made in this paper of the public health significance of 131I discharges from BWR nuclear power plants. The in-plant barriers to the transport of 131I through a BWR nuclear power plant to the site environs were found to function at a high efficiency. As a result, the critical phase of the environmental surveillance program for Dresden, a BWR without gaseous holdup, was concluded to be the submersion dose from noble gases, not 131I in milk. For a BWR nuclear power plant employing gaseous holdup, it was concluded that the ingestion 131I dose may be more limiting than the submersion dose from noble gases.