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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

The Development of a Demonstration Passive System Reliability Assessment

In this paper, the details of the development of a demonstration problem to assess the reliability of a passive safety system are presented. An advanced small modular reactor (advSMR) design, which is a pool-type sodium fast reactor (SFR) coupled with a passive reactor cavity cooling system (RCCS) is described. The RELAP5-3D models of the advSMR and RCCS that will be used to simulate a long-term station blackout (SBO) accident scenario are presented. Proposed benchmarking techniques for both the reactor and the RCCS are discussed, which includes utilization of experimental results from the Natural convection Shutdown heat removal Test Facility (NSTF) at the Argonne National Laboratory. Details of how mechanistic methods, specifically the Reliability Method for Passive Systems (RMPS) approach, will be utilized to determine passive system reliability are presented. The results of this mechanistic analysis will ultimately be compared to results from dynamic methods in future work. This work is part of an ongoing project at Argonne to demonstrate methodologies for assessing passive system reliability.

Bucknor, Matthew↗
DOE OSTI2021

METHODOLOGY AND APPLICATION OF PHYSICAL SECURITY EFFECTIVENESS BASED ON DYNAMIC FORCE-ON-FORCE MODELING

This paper describes ongoing work within the Light Water Reactor Sustainability (LWRS) Program at Idaho National Laboratory (INL) to optimize security and cost of nuclear power plants (NPPs). It reviews the conservatisms in conventional physical security posture and regulations. It introduces the dynamic risk assessment tool developed at INL, Event Modeling Risk Assessment using Linked Diagrams (EMRALD). The dynamic assessment methodology leverages EMRALD to process results of force-on-force (FOF) simulations and crediting safety mitigation actions from probabilistic risk assessment (PRA) models as well as diverse and flexible coping strategies (FLEX) mitigation strategies. Timing information from these simulations are compared against the available time to perform mitigations obtained from Reactor Excursion and Leak Analysis Program (RELAP5) simulations. To illustrate the methodology, a station blackout (SBO) attack scenario was modeled in commercially available FOF simulation tools. The simulation results provide valuable insights into possible attack outcomes and as the probabilistic risk of a core damage event given these outcomes. Safety mitigation procedures were modeled in EMRALD, and were dependent on the attack outcomes by considering human operator uncertainties. RELAP5 simulations incorporating human and hardware uncertainties were performed to estimate the distribution of time-to-core damage. The results demonstrate that, even in the extreme case of a successful adversarial attack, plant mitigation strategies provide significantly high-likelihood of preventing radiological release. The proposed modeling and simulation framework of integrating FLEX equipment with FOF models enables the NPPs to credit FLEX portable equipment in the plant security posture, resulting in an efficient and optimized physical security.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A defense in depth approach for nuclear power plant accident management

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Accident-tolerant fuel: quantifying relative risks and timescales in severe accidents

For such severe accident scenarios, a key objective for alternative 'accident tolerant' fuels (or ATFs) is to increase the threshold temperature for possible accelerated hydrogen production and/or fuel damage and core degradation. By reducing hydrogen production, this change to more ATF should also give more time for restoring power and cooling, for mitigating actions like emergency equipment deployment, and consequently effect on- and off-site evacuation decisions. There is always a finite and non-zero probability of core damage due to the likelihood and chance of not restoring power and cooling. Using the Fukushima complete Station Black Out (SBO) and Loss-of-Cooling (LOC) scenario as a starting point, this paper establishes a simple limiting estimate for the potential changes in the timescales available for power and cooling recovery and for emergency evacuation decision-making made possible by using ATFs. The straightforward analysis also provides a basis for estimating the realistic probability of avoiding core damage for a range of assumed ATF failure properties or criteria, and the needed backup cooling requirements. The analysis includes the effects on the recovery probability caused by significant disruption and chaos. The results show that for this accident scenario, there is a relatively small benefit or increase in recovery probability due to only having higher ATF failure limits. The largest relative benefit arises from not only having ATF but also assuring longer cooling by extending the boil off or time before core uncover and damage onset, from somehow adding and/or injecting sufficient make up water, and/or rejecting decay heat, say for five or more days. Obviously, as an added benefit, some ATF concepts avoid or reduce hydrogen production, which improves containment integrity but does not significantly alter the core damage or failure probability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Performance Analyses of Passive Containment Cooling System for a Concrete Type Containment

The containment of nuclear power plants is the final barrier to keep the radioactive materials from uncontrolled release to the environment. To keep the containment integrity, pressure and temperature must be controlled by the spray system which relies on active components for PWRs with a concrete containment. To achieve a passive means to keep the containment integrity, therefore, cooling should be achieved by forming a natural circulation loop between internal cooling devices connected to an external heat sink. A system like this is often called Passive Containment Cooling System (PCCS). In the early phase of APR-1400 development, this concept was studied and VVER-1200 is also known to adopt the PCCS. The PCCS consists of a cooling device installed at the inside of the containment which is a tube-bank type heat exchanger where cooling occurs at the exterior of the tubes and a heat sink which is called Passive Containment Cooling Tank (PCCT) located at the outside of the containment on the top of the auxiliary building. In this paper, we introduce performance analysis results, utilizing the GOTHIC computer code. The PCCS performance analysis model was developed for APR-1400 which has a thermal power of 4000 MW and considers PCCS as an option to improve its safety characteristics. The PCCS which is considered in this paper is to cope with both Design Basis and Severe Accidents, considering SBO conditions. The pressure and temperature transients due to the Loss of Coolant Accident (LOCA) were analyzed with and without PCCS together with a normal case using the spray system. And, in particular, limitations of the PCCS in regard to mission time will be discussed. Through this study, it was found that the PCCS being considered as an option to improve the safety of APR-1400 can have almost same capability to control the containment pressure for LBLOCA as the containment spray system does. Although it would depends on the heat removal capacity of a given design such as the size of the heat transfer area, the PCCS could even replace the containment spray system. However, the PCCS may lose its capability after 24 hours as its temperature reaches that of the containment atmosphere. To extend the mission time of PCCS, either cooling or supplementary cold water needs to be provided to the external heat sink of PCCS. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

SIT Makeup Effect on the Loss of Residual Heat Removal during Mid-loop Operation

For the loss of cooling during the mid-loop operation, the plant needs to have a sufficient time for operator action without active makeup measures such as a loss of alternating current (AC) power. This concern is raised after the Fukushima accident. Especially, in addition to the loss of residual heat removal (LORHR), if the AC power or active makeup measures are unavailable, the passive makeup using gravity feed or pressurized feed will provide sufficient time before core uncovery. This paper investigates means of the passive makeup measures using safety injection tank (SIT) after the LORHR event to prevent or delay the core uncovery and fuel failure for the typical advanced power reactor 1400 (APR1400). Unfortunately, the refueling water storage tank of APR1400 cannot be used for gravity makeup since it is located on the bottom of the containment with a lower elevation than the reactor vessel. So, the SIT makeup and the effect on the core uncovery are investigated. The results show that gravity makeup of a SIT provides insufficient means but the slightly pressurized SIT is required for sufficient operator action time for the LORHR with the station black out (SBO) during the mid-loop operation for the APR1400. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analytical Models for Terry Turbine Nozzle

As part of the efforts to understand the unexpected 'self-regulating' mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine should be developed and implemented in reactor system analysis or severe accident codes. The Terry turbine has been used for all the BWR RCIC systems and for PWR AFW (Auxiliary Feed Water) systems. The Terry turbine is essentially a solid cylindrical wheel with multiple machined semi-circular 'buckets' that are shaped into the body of the wheel. Fixed nozzles and reversing chambers surround the wheel inside the turbine casing. High pressure steam is accelerated to supersonic flow inside the turbine nozzle. The kinetic energy is then converted to shaft work by the impulse force on the turbine buckets. A Fully implicit and strongly coupled RCIC system model has been developed in the RELAP-7 code and used for simplified BWR SBO (Station Black-Out) simulations in the past. In this model, a generic turbine model was used to conserve mass and energy while turbine operation characteristic curves were used to obtain non-dimensional mass flow rate and thermal efficiency. This model could be used for simulating RCIC off-design behavior if off-design operation characteristic curves were available. However, No such curves currently exist for the Terry turbine system due to its unique pure impulse design. A set of mechanistic Terry turbine models suitable for system code implementation has recently been developed by Sandia National Labs. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model which was obtained from a large number of CFD simulations. In this paper, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. These sets of analytical models are simple and generic, and suitable for use in system analysis codes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Pressure Drop-Flow Rate Curves for Single Phase Steam in Steam Generator U-tubes During Severe Accidents

Thermally induced steam generator tube rupture (TI-SGTR) can occur in pressurized water reactors (PWRs) with U-tube steam generators (SGs) during high pressure severe accident sequences resulting in containment bypass. Robust natural circulation patterns including in-vessel natural circulation and countercurrent flow to and from the SGs of superheated steam and other hot gases heats cooler structures in the reactor coolant system (RCS). The buoyancy driven flows return cooler and denser gases back to the core slowing the progression of some core degradation phenomena and failure of the lower head. As much as 50 percent of the decay energy and energy produced from core degradation oxidation reactions can be transported from the core to SGs resulting in the heat up and thermal creep of RCS pressure boundary components including hot leg nozzles, pressurizer surge line, and the SG tubes, many of which contain flaws. The creep failure of an individual tube is dependent on the transitory heating of the tube, a function of the mass flow rate through the tube and local temperature history of steam entering the tube. Buoyancy driven natural circulation flow rates are determined by changes in gravitational head, a combination of changes in fluid density due to heat transfer and change in elevation, and irreversible pressure drops resulting primarily from friction loss in the flow paths and other form losses. The nonuniform flow distribution in U-tubes of varying lengths and heights during single-phase liquid and two-phase natural circulation has been extensively studied in scaled test facilities and verified numerically in previous works. We hypothesize the countercurrent natural circulation of single-phase superheated steam in U-tube SGs is governed by the same characteristic pressure drop-flow rates curves of single-phase liquid and two-phase natural circulations in U-tubes connected to common plena derived in. Secondly, we propose flow distribution is a function of U-tube geometry - height, length, hydraulic diameter, etc., and the larger SGs of Combustion Engineering (CE) style 2 x 4 PWRs, which are the focus of this study, will show greater nonuniformity in tube by tube flow rates and ratio of forward flowing tubes to reverse flowing tubes. This work derives characteristic pressure drop-flow rate curves for single-phase superheated steam in CE-type SGs during a high pressure station blackout (SBO) sequence with failed auxiliary feedwater (AFW) and dry secondary side which is closely related to the TI-SGTR event. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Korean Localization Program for Performance Verification Test of APR1400's RCP

The Korean government together with the Korean nuclear society completed the development of technologies for the design, manufacturing, and performance verification test of the APR1400's Reactor Coolant Pump (RCP) as one of the national programs for the localization of deficient nuclear technology. The RCP localization program was driven by KAERI, which was responsible for the development of the performance verification and evaluation technology, and by DOOSAN heavy industrial company, which was in charge of the design and manufacturing of the RCP. The test facility for the localized APR1400's RCP was constructed at the KAERI site at the end of 2012. The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 deg. C, 11.7 m{sup 3}/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, a type test for the new-type of RCP was performed successfully. In 2014 and 2015, eight production tests for the localized RCPs, which will be installed in a commercial nuclear power plant in Korea, were successfully completed. As a final stage of the localization program of the RCP, KAERI will perform a seal LOCA test to verify the leakage characteristics and produce experimental data on the localized seal assembly for a safety analysis during a station black out (SBO) condition. In this summary, several technical issues developed through a series of research processes for a performance verification test of the localized RCP are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Water Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Lastly, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development and Implementation of Mechanistic Terry Turbine Models in RELAP-7 to Simulate RCIC Normal Operation Conditions

As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia’s original work [1], have been developed and implemented in the RELAP-7 code to simulate the RCIC system. In 2016, our effort has been focused on normal working conditions of the RCIC system. More complex off-design conditions will be pursued in later years when more data are available. In the Sandia model, the turbine stator inlet velocity is provided according to a reduced-order model which was obtained from a large number of CFD (computational fluid dynamics) simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine stator inlet. The models include both an adiabatic expansion process inside the nozzle and a free expansion process outside of the nozzle to ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input information for the Terry turbine rotor model. The analytical models for the nozzle were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The newly developed nozzle models and modified turbine rotor model according to the Sandia’s original work have been implemented into RELAP-7, along with the original Sandia Terry turbine model. A new pump model has also been developed and implemented to couple with the Terry turbine model. An input model was developed to test the Terry turbine RCIC system, which generates reasonable results. Both the INL RCIC model and the Sandia RCIC model produce results matching major rated parameters such as the rotational speed, pump torque, and the turbine shaft work for the normal operation condition. The Sandia model is more sensitive to the turbine outlet pressure than the INL model. The next step will be further refining the Terry turbine models by including two-phase flow cases so that off-design conditions can be simulated. The pump model could also be enhanced with the use of the homologous curves.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preliminary Two-Phase Terry Turbine Nozzle Models for RCIC Off-Design Operation Conditions

This report presents the effort to extend the single-phase analytical Terry turbine model to cover two-phase off-design conditions. The work includes: (1) adding well-established two-phase choking models – the Isentropic Homogenous Equilibrium Model (IHEM) and Moody’s model, and (2) theoretical development and implementation of a two-phase nozzle expansion model. The two choking models provide bounding cases for the two-phase choking mass flow rate. The new two-phase Terry turbine model uses the choking models to calculate the mass flow rate, the critical pressure at the nozzle throat, and steam quality. In the divergent stage, we only consider the vapor phase with a similar model for the single-phase case by assuming that the liquid phase would slip along the wall with a much slower speed and will not contribute the impulse on the rotor. We also modify the stagnation conditions according to two-phase choking conditions at the throat and the cross-section areas for steam flow at the nozzle throat and at the nozzle exit. The new two-phase Terry turbine model was benchmarked with the same steam nozzle test as for the single-phase model. Better agreement with the experimental data is observed than from the single-phase model. We also repeated the Terry turbine nozzle benchmark work against the Sandia CFD simulation results with the two-phase model for the pure steam inlet nozzle case. The RCIC start-up tests were simulated and compared with the single-phase model. Similar results are obtained. Finally, we designed a new RCIC system test case to simulate the self-regulated Terry turbine behavior observed in Fukushima accidents. In this test, a period inlet condition for the steam quality varying from 1 to 0 is applied. For the high quality inlet period, the RCIC system behaves just like the normal operation condition with a high pump injection flow rate and a nominal steam release rate through the turbine, with the net addition of water to the primary system; for the low quality inlet period, the RCIC turbine shaft work dramatically decreases and results in a much reduced pump injection flow rate, and the mixture flow rate through the turbine increases due to the high liquid phase flow rate. The net effect for this period is net removal of coolant from the primary loop. With the periodic addition and removal of coolant to the primary loop, the self-regulation mode of the RCIC system can be maintained for a quite long time. Both the IHEM and Moody’s models generate similar phenomena; however noticeable differences can be observed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of the concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.

36 MATERIALS SCIENCE↗
DOE OSTI2014

The RISMC Approach to Perform Advanced PRA Analyses

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these two factors on the safety of the plant, the RISMC Pathway aims to develop simulation-based tools and methods to assess risks for existing nuclear power plants in order to optimize safety. This pathway, by developing new methods, is extending the state-of-the-practice methods that have been traditionally based on logic structures such as Event-Trees and Fault-Trees. These static types of models mimic system response in an inductive and deductive way respectively, yet are restrictive in the ways they can represent spatial and temporal constructs. RISMC analyses are performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a boiling water reactor system during a station blackout accident scenario. We employ the system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Our analysis is in fact performed by: 1) sampling values of a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the set of simulation runs. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station blackout accident scenarios. We are able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

97 MATHEMATICS AND COMPUTING↗