Commercial Nuclear Power Plants Research Articles

MELCOR severe accident analysis for a natural circulation small modular reactor

A natural circulation small modular reactor (NCSMR) is designed and the severe accident of the NCSMR was analyzed with MELCOR code in this work. The NCSMR is a 330 MWt light water reactor no reactor coolant pumps and no active safety injection systems similar to the design of NuScale reactor. A MELCOR code model was developed for the reactor. The MELCOR model shows consistent results in steady state operation with the design values. The NCSMR severe accident analyzed in this work was induced by the small break loss of coolant accident in the reactor pressure vessel and a crack in the containment pool wall. The decay heat removal system is assumed to fail for the main steam line break. The results indicated that the RPV pressure decreases quickly while the containment pressure increases followed by the opening of reactor vent valves. The core exposes and heat up with the coolant leakage and finally collapsed to the lower plenum, leading to the lower head breach. High temperature core corium quickly broke the relatively thin containment vessel wall, and was discharged to the containment pool floor, leading to the molten corium concrete interaction. The fission products release during the accident process was also analyzed. The severe accident sequence shows that the NCSMR has much longer time duration for the reactor vessel lower head failure to occur compared to conventional commercial nuclear power plant under severe accident conditions. The analysis results help identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents in NCSMR.

Integrated Blast Resistance Model of Nuclear Power Plant Auxiliary Facilities

Standards, guidelines, manuals, and researches refer mainly to the required protection of a nuclear power plant (NPP) containment structure (where the reactor's vessel is located) against different internal and external extreme events. However, there is no consideration regarding the man-made extreme event of external explosion resulting from air bomb or cruise missile. A novel integrated blast resistance model (IBRM) of NPP's reinforced concrete (RC) auxiliary facilities due to an external above ground explosion based on two components is suggested. The first is structural dynamic response analysis to the positive phase of an external explosion based on the single degree-of-freedom (SDOF) method combined with spalling and breaching empirical correlations. The second is in-structure shock analysis, resulting from direct-induced ground shock and air-induced ground shock. As a case study, the resistance of Westinghouse commercial NPP AP1000 control room, including a representative equipment, to an external above ground blast loading of Scud B-100 missile at various standoff distances ranging from 250 m (far range) till contact, was analyzed. The structure's damage level is based on its front wall supports' angle of rotation and the ductility ratio (dynamic versus elastic midspan displacement ratio). Due to the lack of specific structural damage demands and equipment's dynamic capacities, common protective structures standards and manuals are used while requiring that no spalling or breaching shall occur in the control room while it remains in the elastic regime. The engineering systems and equipments' spectral motions should be less than their capacity. The integrated blast resistance model (IBRM) of the structure and its equipment may be used in wider researches concerning other NPP's auxiliary facilities and systems based upon their specifications.

Applying and adapting the Swedish regulatory system for decommissioning to nuclear power reactors – The regulator's perspective

Half of the original 13 Swedish nuclear power reactors will be shut down by 2020. The decommissioning of these reactors is a challenge for all parties involved, including the licensees, the waste management system, the financing system, and the Swedish Radiation Safety Authority (SSM). This paper presents an overview of the Swedish regulations for decommissioning of nuclear facilities. It describes some of the experiences that SSM has gained from the application of these regulations. The focus of the present paper is on administrative aspects of decommissioning, such as SSM's guidelines, the definition of fundamental concepts in the regulatory framework, and a proposed revision of the licensing process according to the Environmental Act. These improvements will help to streamline the administration of the commercial nuclear power plant decommissioning projects that are anticipated to commence in Sweden in the near future.

Reactor core transient analysis of an innovative high-level nuclear waste transmuter with metal fuel

For the sustainable development of nuclear energy, reducing high-level nuclear wastes in spent fuels discharged from commercial nuclear power plants is very important. Partitioning and transmutation of minor actinides from these spent fuels are critical for the effective disposal of high-level wastes. However, transmutation in current operational reactors is not efficient and brings some safety problems. Therefore, a dedicated accelerator-driven subcritical transmuter named highly efficient industrial transmuter (HEIT) was newly proposed to solve the problems. This system utilizes the uranium-free metallic dispersion fuel and is of high power density. This paper focuses on the transient analysis of HEIT to prove that it is feasible for the future nuclear industry. Temperatures, cladding stresses, and cumulative creep damage fractions are analyzed. Furthermore, the burnup dependence is also investigated. Three typical transients including the unprotected loss of flow, the beam overpower (BOP), and the unprotected transient overpower are evaluated. Numerical results conclude that in the reactivity insertion transient and the BOP transient, this new system exhibits excellently in safety. However, the shut-down system is required in flow loss case which lasts more than half an hour. The safety margin changes with core depletion. For unprotected loss of flow, the end of lifetime case is the most important one to care. While for BOP and unprotected transient overpower, both the beginning and end of lifetime should be accommodated. The conclusions will be useful in the future design of HEIT and similar systems. Copyright © 2017 John Wiley & Sons, Ltd.

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1×10−7mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1mSv/yr. However, in the drinking well scenario, the maximum exposure dose for engineered vault type disposal facility was assessed as 2.022mSv/yr where the value exceeded the regulatory limit of 1mSv/yr. The maximum exposure dose for rock-cavern type disposal facility was calculated to be 0.634mSv/yr, whose results was relatively very close to the regulatory limit considering high uncertainty of long-term environmental conditions. It was demonstrated that DSRSs including the radionuclides of 14C, 226Ra and 241Am significantly contributed to the large portion of exposure dose to the public based on the long-term safety assessment. Therefore, it was recommended that the near surface disposal of DSRSs containing 14C, 226Ra and 241Am should be restricted and managed by long-term interim storage option in order to minimize their potential radiological health effects.

Development of Passive In-Core Cooling System for Nuclear Safety Using Hybrid Heat Pipe

The importance of passive safety for commercial nuclear power plants has been emphasized after the nuclear accidents that occurred at Three Mile Island and Fukushima. A combination of unexpected human errors, severe natural disasters, and defects of system designs led to the accidents, thereby highlighting the vulnerability of established safety systems of commercial nuclear power plants. Various passive safety systems are under development to mitigate design-basis accidents. However, several uncertainties and problems have been pointed out. As a solution to the problems, this paper proposes various designs for a passive in-core cooling system (PINCs) based on hybrid heat pipes. The feasibility and coolability of the PINCs as a passive safety system for commercial pressurized water reactors was investigated using experimental works and numerical analyses. The PINCs showed sufficient coolability to mitigate station blackout conditions by delaying core uncovery. Additionally, several PINCs concepts for advanced nuclear power plants such as a small modular reactor and Generation IV reactors are suggested.

Comparison of fast neutron rates for the NEOS experiment

The fast neutron rate is measured at the site of NEOS experiment, a short baseline neutrino experiment located in a tendon gallery of a commercial nuclear power plant, using a 0.78-liter liquid scintillator detector. A pulse shape discrimination technique is used to identify neutron signals. The measurements are performed during the nuclear reactor-on and off periods and found to be ~20 per day for both periods. The fast neutron rate is also measured at an overground site with a negligible overburden and is found to be ~100 times higher than that at the NEOS experiment site.

Generation III pressurized water reactors and China’s nuclear power

The design philosophy, overall performance, safety, and economy of three typical generation III (GIII) pressurized water reactors, EPR, AES2006, and CAP1400, are analyzed comprehensively in this paper. Based on comparison with and the lessons learned from the Fukushima nuclear accident, we forecast a future reactor for China’s commercial nuclear power plant. Moreover, we put forward important technological fields of GIII nuclear power plants to which attention should be paid, including the enhancement of defense in depth, defense against extreme external events, severe accident mitigation, design simplification and standardization, improvement in economic competitiveness, load following capability, and adaptation to climate change.

Toward Resilience to Nuclear Accidents: Financing Nuclear Liabilities via Catastrophe Risk Bonds

In light of the 2011 Fukushima disaster, recent discussion has focused on finding the best nuclear storage options, maximizing the oversight power of global institutions, and strengthening safety measures. In addition to these, the development of dependable liability coverage that can be tapped in an emergency is also needed and should be considered thoughtfully. To succeed, financing is essential using special-purpose instruments from the global bond market, which is as big as US$175 trillion. Thus, in this paper, for the first time, a two-coverage-type trigger nuclear catastrophe (N-CAT) risk bond for potentially supplementing the covering of U.S. commercial nuclear power plants (NPPs) beyond the coverage per the Price Anderson Act as amended, and potentially other plants are proposed and designed worldwide. The N-CAT peril is categorized by three risk layers: incident, accident, and major accident. The pricing formula is derived by using a semi-Markovian dependence structure in continuous time. A numerical application illustrates the main findings of the paper.

Geological Disposal of Nuclear Waste: a Primer

The back-end of the nuclear fuel cycle has become the Achilles Heel of nuclear power. After more than 50 years of effort, there are, at present, no operating nuclear waste repositories for the spent nuclear fuel from commercial nuclear power plants or for the high-level waste from the reprocessing of spent fuel. The articles in this issue of Elements describe the status of geological disposal in salt, crystalline rock, clay, and tuff, as presently developed in five countries.

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of material degradation mechanisms and on effective strategies to manage the risks of debris-induced failures during a LOCA.

Long-stalled U.S. nuclear plant set to come on-line

For the first time in 20 years, a new commercial nuclear power plant will come on-line in the U.S. when Watts Bar Unit 2 in southeastern Tennessee begins operations later this summer. The plant will be the first to meet new regulations from the Nuclear Regulatory Commission after the 2011 earthquake and tsunami that damaged reactors at the Fukushima Daiichi nuclear power plant in Japan, the Energy Information Administration (EIA) says. After a decades-long delay, construction of the plant, which began in 1973, concluded in 2015 at a total cost of $4.7 billion. The plant is expected to add 1,150 MW of electricity generating capacity to the area. Together with Watts Bar Unit 1, which began operating in 1996, Watts Bar Unit 2 will generate enough electricity to supply 1.3 million homes daily, according to Tennessee Valley Authority, a federally owned electric utility. EIA says that within the next four

National Academies recommend changes at U.S. plants

Damage to spent nuclear fuel storage facilities at the Fukushima Daiichi nuclear power plant should be a wake-up call to plant operators and regulators in the U.S., says a report from the National Academies of Sciences, Engineering & Medicine. The report recommends strengthened and redundant surveillance to improve plant operators’ ability “to measure real-time conditions in spent fuel pools and maintain adequate cooling of stored spent fuel during severe accidents and terrorist attacks.” Reactors at Japan’s Fukushima Daiichi nuclear power plant sustained severe damage after a 9.0 magnitude earthquake and the tsunami that followed it in March 2011. In the wake of the incident, the National Academies initiated a congressionally mandated study of the accident to learn how to improve the safety and security of more than 60 U.S. commercial nuclear power plants. In its second and final report on the issue, the National Academies found that a lack of

Soldered joints—an essential component of demountable high temperature superconducting fusion magnets

Demountable superconducting magnet coils would offer significant benefits to commercial nuclear fusion power plants. Whether large pressed joints or large soldered joints provide the solution for demountable fusion magnets, a critical component or building block for both will be the many, smaller-scale joints that enable the supercurrent to leave the superconducting layer, cross the superconducting tape and pass into the solder that lies between the tape and the conductor that eventually provides one of the demountable surfaces. This paper considers the electrical and thermal properties of this essential component part of demountable high temperature superconducting (HTS) joints by considering the fabrication and properties of jointed HTSs consisting of a thin layer of solder (In52Sn48 or Pb38Sn62) sandwiched between two rare-earth-Ba2Cu3O7 (REBCO) second generation HTS coated conductors (CCs). The HTS joints are analysed using numerical modelling, critical current and resistivity measurements on the joints from 300 to 4.2 K in applied magnetic fields up to 12 T, as well as scanning electron microscopy studies. Our results show that the copper/silver layers significantly reduce the heating in the joints to less than a few hundred mK. When the REBCO alone is superconducting, the joint resistivity (RJ) predominantly has two sources, the solder layer and an interfacial resistivity at the REBCO/silver interface (∼25 nΩ cm2) in the as-supplied CCs which together have a very weak magnetoresistance in fields up to 12 T. We achieved excellent reproducibility in the RJ of the In52Sn48 soldered joints of better than 10% at temperatures below Tc of the REBCO layer which can be compared to variations of more than two orders of magnitude in the literature. We also show that demountable joints in fusion energy magnets are viable and need only add a few percent to the total cryogenic cost for a fusion tokamak.

Progress of Experimental Research on Nuclear Safety in NPIC

Abstract Two kinds of Generation III commercial nuclear power plants have been developed in CNNC (China National Nuclear Corporation), one is a small modular reactor ACP100 having an equivalent electric power 100 MW, and the other is HPR1000 (once named ACP1000) having an equivalent electric power 1 000 MW. Both NPPs widely adopted the design philosophy of advanced passive safety systems and considered the lessons from Fukushima Daichi nuclear accident. As the backbone of the R&D of ACP100 and HPR1000, NPIC (Nuclear power Institute of China) has finished the engineering verification test of main safety systems, including passive residual heat removal experiments, reactor cavity injection experiments, hydrogen combustion experiments, and passive autocatalytic recombiner experiments. Above experimental work conducted in NPIC and further research plan of nuclear safety are introduced in this paper.

Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate

Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical assembly has been carried out in neutronic, thermal hydraulics and safety analysis. Ten benchmark critical core configurations of critical assembly are established and investigated to show safety during normal opeartion and accident conditions. Design calculation results show that NPP fuels are fully adequate for critical assembly operating under nominal power 100W and having average neutron flux about 3×109 neutron/cm2.s.

Moderator design studies for a new neutron reference source based on the D–T fusion reaction

Abstract The radioactive isotope Californium-252 (252Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D2O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuteriumtritium (D–T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252Cf. To be viable, the 14 MeV D–T output spectrum must be significantly moderated to approximate common workplace environments. This paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2–5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered.

Modeling and simulation of radio-iodine released inside the containment as result of an accident

Significant amount of radioactive iodine could release from a commercial Nuclear Power Plant (NPP) in case of accident or due to rapture in primary loop piping. Quantification of iodine inside the containment is essential to study during normal or emergency situations of a NPP because it delivers the thyroid and lungs dose to workers in control room and in the surroundings. In this work, a semi kinetic model Iodine Activity Released in Containment air (IARIC) has been developed for parametric sensitivity study of iodine inside the containment. The model has been developed and implemented in FORTRAN 90 which contains deterministic and as well as a kinetic approach. The IARIC uses the isotope inventory evaluated by ORIGEN2 code (Croft, A. G., 1980) and isotopes data to determine the quantification of airborne iodine and its isotopes as a function of time. The sensitivity analysis of airborne iodine has been carried out as time dependent functions of recirculation rate, leakage rate, release rate with coolant, time dependent mixing rate in coolant. The effect of spray flow rate and spray droplet size has also been investigated. Kinetic simulation with continuous source of iodine as time dependent function of mixing rate (wx) shows that wx has a significant effect on airborne iodine activity. The iodine spray cleanup rate evaluated by IARIC are 3.22 h−1 and 6.68 h−1 which are in consistent with the rate found in literature.

Post-test analysis of the ROSA/LSTF and PKL counterpart test

Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety.Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations.

Entry into force of the Convention on Supplementary Compensation for Nuclear Damage

There are 431 commercial nuclear power plants around the world.1 On 14 April 2015, 193 of these power plants were covered by a nuclear liability instrument (118 power plants by the Paris Convention2 and 75 by the Vienna Convention3). With the entry into force of the Convention on Supplementary Compensation for Nuclear Damage (CSC)4 on 15 April 2015, the number of power plants covered by a nuclear liability