Westinghouse Electric Company Research Articles

A thermogravimetric comparative study of the high temperature steam oxidation of bare and pre-oxidized Zircaloy-4, M5Framatome and Optimized ZIRLO™

For Loss of Coolant Accident (LOCA) analysis, high-temperature oxidation kinetics in steam of Zircaloy-4 and the two Nb-bearing alloys used in French Pressurized Water Reactors (PWRs), M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. and Optimized ZIRLO™22Optimized ZIRLO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners., have been studied using Thermo-Gravimetric Analysis (TGA) technique in the temperature range of 900–1200 °C. A specific test procedure allowing the use of a high steam flow rate and good steam supply to the whole specimen surface, as well as fast heating of the specimen, has been developed. Clean kinetic data, not affected by steam starvation, were obtained on the bare, as-received alloys. For as-received claddings, oxidation kinetics are comparable and close to parabolic for the three alloys at 1100 °C and above. Below 1100 °C, a significant deviation from the parabolic kinetic regime was observed, and M5 Framatome oxidizes slower than Zircaloy-4 and Optimized ZIRLO. Thanks to the fast-heating protocol, the influence of pre-oxidation, simulating corrosion in normal reactor operation, could be investigated, avoiding evolution of the pre-oxidation scale during a slow, long heating phase. Pre-oxidized Zircaloy-4 and M5 Framatome were investigated. A strong protective effect is observed below 1200 °C, while at 1200 °C, the pre-oxidation layer appears to be much less protective, possibly due to the monoclinic to tetragonal zirconia phase transformation.

The Passive Heat Removal Facility: Testing and Demonstrating an Innovative Decay Heat Removal System

Innovative reactor concepts are being studied by several research institutions and private entities for their role against climate change and energy poverty. The Generation IV International Forum, committed for more than 20 years in support of advanced reactors, has drawn up a series of objectives for the new class of nuclear plants, among which is an improvement in safety and economy through passive safety systems for the removal of decay heat. One of the most studied technologies is the lead fast reactor (LFR), whose coolant has a high boiling point, excellent shielding capabilities, and good heat transfer in forced and natural circulation, as well as neutron properties suitable for a hard spectrum. These allow for designing systems with high simplification and wide safety margins. On the other hand, opacity, corrosion, and innovative design choices require demonstration of the technology in experimental campaigns before using it in the industrial field. In recent years, Westinghouse Electric Company has begun the conceptual study of an intermediate size LFR [~460 MW(electric)] which, by exploiting the opportunities of the technology, aims at marketability over the next decade. One of the features is a passive heat removal system that allows, through different heat exchange mechanisms including conduction, convection, and radiation, for the transfer of decay heat from the reactor block to a pool of water inside the containment. The system is designed for indefinite heat removal thanks to channels that allow outside air to replace water following complete boiling. For the geometry of the system, the size, the materials, and the heat transfer mechanisms, an experimental activity is required to validate the prediction of the calculation codes and potentially support design optimization. The UK Department for Business, Energy & Industrial Strategy has recently subsidized the design, procurement, installation, and operation of the Passive Heat Removal Facility, an experimental facility to study the innovative safety system. This paper presents the activities of scaling, design, pretesting, and installation of that facility.

PROTECTIVE VACUUM-ARC COATINGS ON ZIRCONIUM ALLOY FUEL CLADDING TO PREVENT CATASTROPHIC ACCIDENTS AT NUCLEAR REACTORS

The processes of vacuum arc deposition of multicomponent coatings based on Cr, Ti, Al, Fe, Y, Si, and their nitrides with a thickness of 5…9 µm for the protection of zirconium alloy fuel rods were developed. The composition, structure, and mechanical properties of the coatings were determined by electron microscopy, energy dispersive X-ray spectroscopy, X-ray diffraction analysis, and nanoindentation. Comparative studies of the protective properties of coatings during high-temperature oxidation of Zr1Nb tubes were carried out. It was found that metal coatings with a high chromium content have excellent protective properties during tests under simulated accident conditions. They are simultaneously resistant to high-temperature corrosion in air at 750…1100 °C and in water steam during thermal cycling up to 1020 °C. Hard nitride coatings, which significantly strengthen the surface of the zirconium alloy and provide good protection against oxidation in air, are less effective in water steam under thermal cycling. The coatings and processes of synthesis developed at NSC KIPT can be useful for improving fuel cladding made of zirconium alloys of the world's leading nuclear fuel producer Westinghouse Electric Company.

The Versatile Loop Facility: A New Infrastructure for Testing Components and Systems of Lead Fast Reactor Technology

There has recently been growing interest in the development of innovative nuclear technologies that offer greater sustainability and cost effectiveness of electricity production. One of the most promising options is the lead fast reactor (LFR) technology. Lead stands out for its favorable neutron properties, allowing a hard neutron spectrum core as well as good shielding, heat transfer, and radioisotope retention capabilities. As lead has a boiling point in excess of 1700°C and does not react exothermically with either air or water, it also allows for the design of a low-pressure reactor block without an intermediate cooling circuit, which is used in other advanced reactor technologies for protecting against the interaction between primary and power conversion system coolants. The deployment of a new fleet of fast reactors is conditional on the control/prevention of the corrosion and erosion effects of the coolant against the structural materials, the systematic characterization of the interaction phenomena between the coolant and fuel and water, and the experimental qualification of innovative systems and components. To support LFR technology development, the UK Department for Business, Energy & Industrial Strategy has recently allocated 10 M£ to a team composed of Westinghouse Electric Company LLC, the Ansaldo Nuclear Group, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development, the University of Manchester, and other organizations for the design, construction, and first operation of a network of eight test infrastructures widespread in the United Kingdom to address the LFR’s highest priority research and development needs. One of the experimental rigs is the Versatile Loop Facility (VLF) currently under construction at the Ansaldo Nuclear Group’s workshop in Wolverhampton, United Kingdom. The plant consists of a lead loop operable up to 650°C and equipped with a 500-kW electric fuel bundle simulator (resembling the Westinghouse LFR bundle) and a hybrid microchannel-type diffusion-bonded heat exchanger (which simulates the primary heat exchanger adopted in the Westinghouse LFR design). The heat removal is delegated to a supercritical water-cooling loop having a design pressure of 330 bar and maximum operating temperatures up to 620°C. In this paper we present the design of the VLF with specific details about its prototypical components and an insight into the construction and installation phases currently underway.

Analysis and Preliminary Design of Primary Heat Exchanger Failure Testing Facility for Lead-Cooled Fast Reactors

This paper presents an assessment aimed at evaluating primary heat exchanger (PHE) failure of the Westinghouse Electric Company Lead-cooled Fast Reactor (LFR) and at designing a facility for testing phenomena involved in such failure. The system thermal-hydraulic code RELAP5/MOD3.3 was used to develop a transient analysis simulation at reactor scale. Because of RELAP5/MOD3.3’s inability to mix working fluids, the steam injection effect was evaluated using the SIMMER-III code. The limits and strengths of both codes are highlighted throughout the paper. The reactor-scale steady-state results are in good agreement with the nominal operating condition. The transient results show that lead pool surface level variation and primary system pressurization during the PHE failure event are limited. The PHE failure testing facility was characterized, and a preliminary layout was developed. A separate-effects transient inside the vessel was analyzed with SIMMER-III and RELAP5/MOD3.3 runs. The simulation outcomes have provided useful data to inform subsequent design stages for the test facility. Different configurations of the facility have been assessed, highlighting the strengths and weaknesses of each design. The most important issue was identified to be lead pool swelling, reaching the vessel’s lid and blocking the pressure relief vent. This poses a safety hazard that must be addressed and has been raised for resolution in subsequent design stages. The so-called V4 configuration is suggested as a starting point for further improvement of the facility. Furthermore, a smaller failure opening and lower lead level in the vessel are suggested.

LES simulation on heavy liquid metal flow in a bare rod bundle for assessment of turbulent Prandtl number

The Westinghouse Electric Company, together with an international team, is developing its next-generation high-capacity nuclear power plant based on lead-cooled fast reactor technology. In general, the Prandtl number of a heavy metal fluid is two orders of magnitude lower than that of water. Because of the low Prandtl number of lead, the thermal boundary layer in the temperature field is much thicker than the viscous boundary layer in the velocity field. Existing engineering turbulence models all use the Reynolds analogy for coupling temperature and velocity fields, but it is not valid for Heavy Lead Metal flow. Therefore, the selection of an appropriate turbulent Prandtl number (Prt) is crucial for lead fluid simulation. The conventional choice of Prt as a constant value of 0.9 is not valid for low-Prandtl-number fluid flow. This paper aims at using an advanced high-fidelity numerical approach, Large Eddy Simulation (LES), to provide detailed insight into the physics of the flow and the associated heat transfer in a rod bundle with liquid metal flows. The results for the global Nusselt number with different Prt values (0.9 ∼ 3.0) indicate that a higher Prt can reduce the predicted Nusselt number by up to 44.1 %. The detailed temperature distributions obtained with RANS and LES are compared to better understand the deviation introduced by the turbulence model. For turbulent flow, the correlations for Nusselt number proposed by Cheng and Tak (2006) are in good agreement with the simulation, especially for a Peclet number < 1000. The analysis shows that the RANS model with Prt = 1.5 shows better agreement on the prediction of local temperature distribution and global Nusselt number. Although the modified local turbulent Prandtl number model shows some improvement over the constant Prt model, the development of a more sophisticated local turbulent Prandtl number model is necessary for further applications.

FEATURES AND DEVELOPMENT PROSPECTS OF THE ODESSA REGION IN THE ENERGY SYSTEM OF UKRAINE

In the modern world, energy is the basis of the development of basic industries, which determines the progress of public production. The growing need to use electricity poses problems in finding sources of its generation. The main problem of choosing sources of electricity production is the impact of energy on the environment. Many countries focus their energy development strategy on renewable energy sources (RES) (solar, wind, bio and others). However, due to their specificity, renewable energy sources cannot ensure the production of the required amount of electricity. The key problem of renewable energy sources is to ensure an uninterrupted supply of energy to end users. In the work, a comparative analysis of the state of the structure of energy resources and their availability in the southern region of Ukraine was carried out. The problem of ensuring energy and environmental security of the Odesa region in conditions of a shortage of traditional energy resources is considered. The current state and prospects of the use of renewable energy sources are considered. Using the method of expert evaluations to justify the choice of the type of acceptable energy source and increase the relative energy security of individual territories, the use of nuclear energy is substantiated. The prospects of placing new power units with the AP1000 reactor unit developed by the Westinghouse Electric Company on the territory of the former construction site of the Odesa Nuclear Thermal Power Plant (Teplodar) were considered. The technical characteristics of the AP 1000 project are given. The obtained results can be the basis of objective assessments regarding the prospects of using nuclear energy to increase the energy security of the Odesa region. The relevance of the work is determined by the importance of international cooperation in atomic energy, which is a strategically significant, innovation-intensive branch of the modern world economy in the conditions of development of the Odesa region.

Development of Conceptual Lead Cartridge Design to Perform Irradiation Experiments in VTR

This paper seeks to introduce the latest design of the Extended Length Test Assembly–Cartridge Lead (ELTA-CL) with associated thermal-hydraulic (TH) assessment and related experiment activities to support the critical component development performed by the ELTA-CL team (Los Alamos National Laboratory, Westinghouse Electric Company, and the University of New Mexico). The goal of the ELTA-CL program is to develop and validate an experimental capability to perform irradiation experiments in the Versatile Test Reactor (VTR) addressing Lead Fast Reactor (LFR) technology gaps, in support of the commercial development of advanced lead-cooled fast reactor concepts. Through a design maturation process and parametric study, a conceptual design is proposed to meet the requirements for material and corrosion testing. Thermal-hydraulic characteristics for the conceptual design at desired operating conditions are assessed with systems-level (one-dimensional) and computational fluid dynamics (three-dimensional) simulations. Along with the conceptual design work, experimental activities for the development of critical components such as the pump and flowmeter are undertaken. From both the modeling study and the experimental results, the design requirements of the Phase 1 ELTA-CL (e.g., 500°C and 2 m/s) are achievable with the current conceptual design. Additional design improvements and safety assessments at both steady-state and transient conditions for the final ELTA-CL design will be pursued.

The speciation of niobium in the oxide layer of an irradiated Low-Tin ZIRLO nuclear material

In this study the oxidation state of niobium, within the oxide layer of a low-tin ZIRLO11ZIRLO is the trade name of a zirconium based alloy produced by Westinghouse Electric Company. irradiated in a nuclear reactor, is examined using synchrotron. The unique set up allows simultaneous acquisition of X-ray florescence (XRF) and X-ray diffraction (XRD) maps. A gradual oxidation and evolution of niobium is observed and quantified which is then compared with the information on the distribution of oxide phases revealed by XRD. The results are discussed with the aim to better understand the hydrogen uptake mechanism for this material, particularly the cause of the increased resistance to hydrogen uptake seen in niobium containing alloys.

Evaluation of multi-cycle physics and economics considering end cycle optimization for a PWR

Nowadays, license renewal of nuclear power plants is difficult due to the political, technical, and economical reasons, and the optimum end-cycle strategy has become an interesting and important research topic. In this paper, commercial cycle optimization strategies considering permanent shutdown have been studied for a Westinghouse Electric Company type PWR core. Both the conventional equilibrium reload cores and end-cycle cores for permanent shutdown are considered for comparison. Multi-cycle optimizations are mainly focused on the economics and neutronics analysis based on the commercial design criteria. In this work, all the neutronic designs and analyses are done with a licensed design code system. A new 18- and 24-month hybrid approach is also introduced to mitigate design challenges for equilibrium reload cores. For the end-cycle optimization, various representativeoperational scenarios and fuel management schemes have been devised and analyzed for multi-cycle optimizations. For each end-cycle operational scenario, detailed multi-cycle designs are performed and an annual total cost is evaluated to find the optimal one. In addition, a sensitivity study for the cost components are performed.

Special Section: ICONE-26

This Special Section in this issue of the ASME Journal of Nuclear Engineering and Radiation Science is dedicated to selected papers from the 26th International Conference on Nuclear Engineering (ICONE-26) held in London, United Kingdom, during July 22–26, 2018. The selected ICONE-26 papers were revised and reviewed to meet the journal standards and requirements.The ICONE conference is a premier global annual conference on nuclear engineering, science, and technology that has a quarter century history and is organized by the Nuclear Engineering Division (NED) of the ASME, the Japan Society of Mechanical Engineers (JSME), and the Chinese Nuclear Society (CNS). The first conference, ICONE-1, was held in 1991 in Tokyo with the JSME and ASME as sponsors. In 2005, the Chinese Nuclear Society (CNS) formally joined as the third sponsor and hosted ICONE-13 in Beijing. The ICONE-series of conferences are global significant events of choice for nuclear professionals to stay technically current, follow industry trends, promote professional development, and provide a platform for information exchange and dissemination. The contributions from industry, government, and academia lead to the success of ICONE as the quality and number of the technical presentations steadily increased. In addition to the archival paper publications, ICONE assists the nuclear industry in developing future generations of nuclear professionals to meet industry needs. This is achieved through an excellent student paper competition program that includes a significant number of students from around the globe, supported and partially funded by the conference sponsors (60 finalists in total consisting of 15 students from each of the following four regions: North America, Europe, China, and Japan with Asia at large).Based on the turnout and the comments received from many of the nearly 900 registered participants, the 26th International Conference on Nuclear Engineering (ICONE-26), held in London, UK, was a great success. In total, 57% of the participants were from academia, 29% from the industry, and 11% from government. Geographically, the largest part of participants came from China (45%), European Union (20%, including 7% from UK), Japan (16%), USA and Canada (11%), and South Korea (6%). More than 670 final technical papers and 760 presentations were made during the ICONE-26 conference from countries worldwide in a total of 15 technical tracks. These statistics show emphatically that ICONE presents a truly worldwide view on Nuclear Power, and there is no better venue to express viewpoints or share information with likeminded professionals.The ICONE-26 organizers invited the authors to submit their revised and augmented papers for journal publication consideration. The selected final papers presented in this special section cover a wide area of topics in nuclear science and technology. Reflecting the state of the nuclear industry, there are several papers dealing with analysis, simulation, management, experiments, context evaluation, and consequences of postulated design basis and severe accidents. This special issue also contains papers related to research and development challenges into advanced reactors, fusion technology, decontamination and decommissioning, radiation protection, and other related topics.The guest editors want to warmly thank the authors and reviewers for their efforts and support in producing this journal issue. During the editorial process, the papers went through the rigorous peer-review process. The ICONE-based special issue or section of the Journal of Nuclear Engineering and Radiation Science is now already traditional and we hope to continue the journal publication of selected augmented papers from the future ICONE conferences. The guest editors would like to extend their thanks to the editorial board of the journal and, in particular, the editor, Dr. Igor Pioro, for their support during the preparation of this special section. Finally, we hope that the readers enjoy this special ICONE-26 based section.Guoqiang Wang, Ph.D. ICONE-26 Conference Chair Fellow ASME Westinghouse Electric Company LLC, Pittsburgh, PA 15235 e-mail: WangG@Westinghouse.comShripad Revankar, Ph.D. ICONE-26 Technical Program Chair Fellow ASME School of Nuclear Engineering,Purdue University,West Lafayette, IN 47907e-mail: shripad@purdue.eduLeon Cizelj, Ph.D. ICONE-26 Conference Co-Chair Fellow ASME Reactor Engineering Division, Jožef Stefan Institute, Ljubljana 1000, Slovenia e-mail: Leon.Cizelj@ijs.siRobert Stakenborghs ICONE-26 Steering Committee Chair Fellow ASME ILD, Inc., Baton Rouge, LA 70808 e-mail: bob@ildpower.comJovica Riznic, Ph.D. ICONE-26 Organizing Committee Co-Chair Fellow ASME Canadian Nuclear Safety Commission, Ottava, ON K1P 59S, Canada e-mail: jovica.riznic@canada.ca

Analysis of experience, safety and prospects of diversification of nuclear fuel at nuclear power plants

The object of research is the design fuel assemblies (FA-A) at nuclear power plants with water-cooled water reactors located in Ukraine. The analysis of the experience and prospects of diversification of design fuel assemblies of pressurized water reactors with alternative heat generating assemblies of Westinghouse Electric Company is carried out. The analysis is carried out on the basis of measures and results of diversification of fuel assemblies at the Temelin nuclear power plant (Czech Republic), as well as the South Ukrainian and Zaporizhzhia nuclear power plants (Ukraine). As a result of the analysis, it is shown that the diversification of design fuel assemblies by alternative heat assemblies by Westinghouse Electric Company provides the necessary nuclear safety conditions for the maximum allowable temperature of the gratings of fuel rods and the temperature of nuclear fuel.The work determines the need for additional analysis of nuclear safety and reliability of the reactor loop equipment in relation to the conditions of critical hydrodynamic shock during the diversification of fuel assemblies. It is revealed that the known results of nuclear safety analysis during diversification of fuel assemblies by traditional accident modeling approaches are not sufficiently substantiated. And they also significantly depend on the negative effects of «differences in the results of accident modeling by different users of the same codes» and «differences in the results of accident modeling by different codes». In addition, well-known deterministic codes do not simulate the conditions and consequences of hydrodynamic shock and various types of thermohydrodynamic instability in the reactor loop. It is shown that it is necessary to develop alternative methods for analyzing nuclear safety and equipment reliability of systems important to safety in diversifying fuel assemblies that are independent of the above negative effects. A calculation analysis of the influence of the coolant speed on the external heat transfer coefficients determined that the safety conditions for the permissible temperature of the Westinghouse Electric Company fuel rods are provided up to a maximum design temperature of 90 °C in safety heat exchangers.

Development of Mechanistic Source Term Analysis Tool SAS4A-FATE for Lead- and Sodium-Cooled Fast Reactors

Fauske & Associates, LLC (FAI), Argonne National Laboratory (ANL), and Westinghouse Electric Company are collaborating within the program “Development of an Integrated Mechanistic Source Term Assessment Capability for Lead- and Sodium-Cooled Fast Reactors.” This program, partially funded by the U.S. Department of Energy through the Gateway for Accelerated Innovation in Nuclear initiative, aims at developing a computational framework for predicting radionuclide release from a broad spectrum of accidents that can be postulated to occur at liquid metal cooled reactor (LMR) facilities. Specifically, the program couples the transient and severe accident analysis capability of the SAS4A/SASSYS-1 code developed by ANL with the radionuclide transport analysis capability of the Facility Flow, Aerosol, Thermal, and Explosion (FATE) code developed by FAI. The testing of both the individual codes and of the coupled system is performed on a generic lead cooled fast reactor (LFR) design that is intended to capture the key differences between the LFR and the sodium fast reactor (SFR), around which the SAS4A/SASSYS-1 code has historically been developed and from which the coupled code inherits some features requiring modification before application to LFR systems. By means of this approach, a computational framework applicable to both LFR and SFR systems will be obtained that will assist LMR developers in performing a realistic, scenario-dependent mechanistic source term (MST) assessment expected not only to strengthen their safety case but also to support easier siting and claims on reduced emergency planning zone requirements. This paper discusses the work being performed to adapt the SAS4A/SASSYS-1 and FATE codes to LFR technology; the code coupling method implemented; and some of the results of the LFR test case, with the latter aimed at demonstrating the progress made toward the development of the MST analysis capability that is ultimately targeted.

Study on Reactor Vessel Air Cooling for Westinghouse Lead Fast Reactor

Westinghouse Electric Company (Westinghouse) is developing its next generation of high-capacity nuclear power plants (NPPs) based on lead fast reactor (LFR) technology: a Generation IV compact, highly simplified, passively safe, and scalable NPP. In addition to superior economics for enabling competitiveness even in the most challenging electricity market, exceptional safety performance is actively pursued in the design of the plant, leveraging the inherent favorable properties of lead coolant as well as safety features intrinsic in the design. Being that decay heat removal (DHR) is an integral part of any NPP’s safety philosophy, a systematic process of concept selection has been employed across a wide variety of DHR system designs. Among them, air cooling outside of the reactor vessel (RV) is one of the concepts that has been actively evaluated by Westinghouse. In this paper, the use of air cooling in nuclear reactors is discussed together with the identification of benefits and challenges associated with RV air cooling in LFR technology. The heat removal capability of this system is assessed with three computer codes, differing in complexity and suitability to “rapid prototyping” design activities carried out by Westinghouse during different phases of plant design. Though the computer codes were developed separately, the results of the three evaluation models tend to support each other, thus increasing confidence in the information provided to progress the Westinghouse LFR design and establish its safety basis. Additional validation through existing and potentially new test data is foreseen as future work within the Westinghouse LFR program.

Unmasking the Fraud at Toshiba

ABSTRACTFollowing its purchase of Westinghouse Electric Company and subsequent macroeconomic events, Toshiba faced declining profits. In response, Toshiba engaged in earnings management through two accounting treatments. First, it delayed the recognition of losses under long-term contracts. Second, it inappropriately applied price masking to account for transfers of components between itself and contract manufacturers. Students using this case will assess how business risks and corporate culture relate to audit risk, and how accounting for price-masking transactions can lead to increased fraud risk. Students will also research aspects of auditing standards related to fraud and accounting estimates. The case is designed for auditing courses and capstone courses with an auditing component.

Development of MELCOR thermal hydraulic model of AP1000 and its verification for a DECL break

The AP1000 is currently an advanced pressurized water reactor with emergency core cooling systems (ECCS) completely dependent on natural circulation. It is under construction in the USA and has been commissioned in China and is being considered by several countries to use this advanced technology for a power source. With this objective in mind, a representative MELCOR model (for the AP1000) has been developed to better understand the thermal hydraulics involved and to assess the performance of the safety systems for a design basis accident such as large break LOCA (LBLOCA) with availability of the safety systems. The results from these calculations were compared to those obtained using WCOBRA-TRAC, developed by Westinghouse Electric Company, which are reported in the AP1000 European Design Control Document.MELCOR was selected for the design assessment activities specifically because of its capabilities to model severe accident scenarios and features such as containment behaviour. It is one of the system thermal hydraulic codes, which are recommended for severe accident analysis in light water reactors. The aim of the current study is to assess the thermal hydraulic models and correlations in MELCOR against a specific code developed for a design basis accident such as a LBLOCA. This exercise will validate the input model for a LBLOCA and then use the input model for station blackout studies and the impact on the containment.A double ended cold leg (DELC) break in the loop containing the Core Makeup Tanks (CMT) is considered for this study. The calculated break flow, peak cladding temperature, accumulator, and CMT injection flow rates are compared to those reported. These calculated results are found to be satisfactory according to ECCS acceptance criteria for the temperature and hydrogen generation. This study has been further extended to analyse the containment behaviour during the accident. The calculations for the transient were extended to observe the behaviour of the Automatic Depressurization System (ADS) and IRWST injection capabilities which can effectively avoid core uncovery for a longer duration of the time. These systems can only work when the reactor system is depressurized. MELCOR calculated a PCT value (occurs just after the break) as 1242 K while the total hydrogen produced was calculated to be 2.216 kg which is only 0.61% of the total potential hydrogen production from oxidation.

Chemical Head Loss Measurements on Multi-Constituent Debris Beds in Small-Scale Experiments

A series of experiments on vertical head loss modules or columns to measure conventional and chemical head loss was carried out to support the resolution of Generic Safety Issue 191 for the Vogtle nuclear power plant (NPP). The head loss (conventional and chemical) was measured on multi-constituent fibrous debris beds of different particulate-to-fiber ratios (η). The debris beds were generated on a horizontal screen following the new procedure developed at the University of New Mexico and are summarized herein. The generated debris beds have been shown to produce repeatable and stable conventional head loss (CHL) and have the ability to detect chemical surrogates. Prototypical Vogtle NPP containment debris materials were used to form three different particulate-to-fiber–ratio (η) debris beds: 6.89 (thin bed), 2 (intermediate bed), and 1.15 (thick bed). The particulates were presented as 90% epoxy paint, 5% inorganic zinc, and 5% latent debris dirt by mass. The obtained results show that the measured CHL increased as the particulate mass increased in the debris beds. The average measured CHL values were 9.37, 6.4, and 5.66 H2O'' for η = 1.15, 2, and 6.89 debris beds, respectively. The debris beds with η = 2 and 1.15 were selected for the chemical head loss experiments.Standard aluminum (Al) chemical precipitates with specific batches were introduced to the head loss columns, and chemical head loss was measured. Precipitates prepared following the WCAP-16530-NP-A procedure [Lane et al., WCAP-16530-NP-A, “Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191,” Westinghouse Electric Company (2008)] or formed in situ by injecting metal salts under two different rates (0.75 and 7.5 mL/min) were tested. The results show that the thin debris bed (~10 mm) was more sensitive to the chemical precipitates prepared following the WCAP procedure compared to the intermediate debris bed (~25 mm) and thick debris bed (~55 mm). The measured chemical head loss was 0.35, 0.1, and 0.02 H2O''/mg of Al filtered by the debris beds. The in situ injection method has shown higher measured chemical head loss per unit mass of filtered precipitates than the WCAP surrogates for the debris beds of η = 2 (intermediate bed) and 1.15 (thick bed). Also, the results show a nonconclusive effect on the injection rate of metal salt to form in situ chemical precipitates on the measured chemical head loss.

Assessing I2S-LWR economic competitiveness using systematic differential capital cost evaluation methodology

One of the main factors impeding nuclear power plant (NPP) construction is their economics. In this study, a systematic differential economics evaluation approach was developed through the use of the Code of Accounts guidelines to assess the costs of nuclear power plants. The methodology, entirely based on publicly available data, may serve as a template to evaluate direct costs for reactors of any size and design at any stage of developments. In particular, this approach was used to assess the costs of the Integral Inherently Safe Light Water Reactor (I2S-LWR).The I2S-LWR is a design concept of a large (∼1000 MWe) light water reactor. One of its key design features promoting inherent safety is implementation of an integral primary circuit configuration, which however necessitates a compact design of the core and primary circuit components.Through the methodology here presented, a representative loop PWR design was taken as a reference and the differential cost was estimated for each individual account based on the design difference, or similarity. Cost scaling techniques were applied to the accounts representing systems that differ from the ones of the reference PWR. Cost estimating techniques were used to evaluate cost of innovative components that are not part of standard PWR designs. By evaluating the cost difference of the I2S-LWR from the standard PWR, rather than the absolute cost, the uncertainty in the estimate is reduced. A similar approach was used by ORNL to estimate the cost of a Fluoride-salt High-temperature Reactor (FHR).A traditional four-loop PWR plant with a core thermal power of 3417 MWth (1144 MWe) was selected as the reference. Costs for that plant were prepared in 1978 by the Department of Energy (DOE) Energy Economics Data Base (EEDB), averaging actual cost incurred in the construction of several nuclear power plants (NPP), itemized with a great level of detail according to the Code of Accounts. This best estimate costs are denoted PWR12-BE. For each account, the cost of equipment, site labor and site material is provided. Industry experts at Westinghouse Electric Company performed a “sanity check” of the cost items, adjusting the cost of several items to match the current market and supply chain data.The detailed cost assessment of I2S-LWR was performed, systematically analyzing cost for each account, and applying the differential economics approach. First, Relative importance of each account, i.e., its contribution to the total cost was established, to help focus analysis on the most significant contributors. Moreover, the accounts describing components that are different than that of the PWR12-BE were identified.The integral configuration of the reactor has important implications on the cost of the reactor plant equipment (accounts 22x). Turbine generator equipment (Accounts 23x) is not believed to be much different than that of the reference design. I2S-LWR structures (Accounts 21x) mainly differ from that of a standard LWR as several buildings (containment building, shield building, annex building, waste processing building, fuel storage building) are integrated into a single building (nuclear island). Yardwork has a higher cost for the I2S-LWR, as the NI is partially below grade. On the other hand, due to its compact Nuclear Island footprint, I2S-LWR facilitates (and includes in its reference version) the use of seismic isolators, which contribute to reducing the direct cost. The total capital investment cost (TCIC), on the $/kW basis, of I2S-LWR is 5.84% lower than that of PWR12-BE, in spite of the lower power output of I2S-LWR. However, for the Western US (0.7 g), benefits of the seismic isolation are more pronounced, and the I2S-LWR total capital investment cost is 13.02% lower than that of PWR12-BE. If the I2S-LWR is compared to a PWR10-BE (traditional loop design, but scaled to the same power level as I2S-LWR), the savings are even higher, in the range 11.12%–17.89%.The analysis indicates that I2S-LWR has potential to offer an economically attractive design, with TCIC lower than that of a nuclear power plant based on a traditional loop PWR design. In other words, I2S-LWR design offers significantly enhanced safety, while at the same time improving economics.The differential economics approach developed in this paper can be used in identifying changes in cost of key components in order to improve the economics of a nuclear reactor design. The method can also help compare the economics of advanced Generation IV reactors and innovative water-cooled SMR (Small Modular Reactors) with respect to standard LWR technologies. In summary, the differential economics approach and can be used as a tool capable of helping stakeholder decisions.

BASIS ACCIDENT ANALYSIS OF A SMALL MODULAR REACTOR

The International Reactor Innovative and Secure (IRIS) is an integrated small modular reactor (SMR) designed by Westinghouse Electric Company and a consortium of organizations. It has a thermal electric power rating of 1000 MWth and an electric power rating of 335 MWe. Being an SMR that features a derivative pressurized water reactor design, it contains the core, upper plenum, lower plenum, pressurizer, reactor coolant pumps, and steam generators in a single reactor vessel. An important feature of the IRIS is passive safety systems. Passive safety systems don’t require mechanical/electrical inputs for operation and rely on natural laws, material properties, and stored internal energy for hazard mitigation. In this work, a RELAP5 model of IRIS was constructed. A small-break loss-of coolant accident (SBLOCA) is simulated where the six major stages of SBLOCA response: (i) break initiation; (ii) blowdown; (iii) vessel depressurization by condensation and automatic depressurization system (ADS); (iv) pressure equalization; (v) reactor vessel/containment vessel depressurization; and (vi) long-term cooling are demonstrated. This paper demonstrates how efficiently the passive cooling systems work in mitigating the hazard.

Verification of AP1000® low-margin PRA sequences based on best-estimate calculations

The Westinghouse AP1000® reactor is an advanced design whose safety systems are mainly passive safety systems. Due to the passive nature of the safety related systems and its dependency on small changes on certain variables (e.g. pressure, friction coefficients) together with the use of a simplified code like MAAP in Probabilistic Risk Assessment (PRA) analyses, it makes necessary to confirm that when core cooling is achieved, thermal-hydraulic (T/H) uncertainties are bounded. The T/H uncertainty evaluation process performed by Westinghouse Electric Company (WEC) identified the low-margin sequences (core uncovery) by expanding PRA Event Trees (ETs). The expanded ETs allowed finding the low-margin risk-important sequences and then a set of low-margin bounding sequences was selected. Then detailed safety analysis methodologies were applied in order to evaluate the bounding sequences demonstrating that the T/H uncertainties were bounded. The Universidad Politécnica de Madrid group has verified the low-margin bounding sequences obtained by WEC with the best-estimate TRACE code in order to verify the previous results and also to study the phenomenology of such sequences through a best-estimate code. This paper presents the results obtained for short-term low-margin bounding sequences. In general, TRACE results do not present important discrepancies with respect to NOTRUMP and WCOBRA/TRAC results although TRACE presents lower values for Peak Cladding Temperature. This analysis has allowed to verify the AP1000 thermal-hydraulic bounding evaluation process performed by WEC for the low-margin risk-important sequences with TRACE code.