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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Thermal-hydraulic validation tests using the SMART-ITL facility with a single train passive safety injection system for the smart design

A test program was launched using a large-scale thermal-hydraulic integral effect test facility of SMART-ITL (or FESTA) to validate the performance of the SMART passive safety system. In this paper thermal-hydraulic validation tests using the SMART-ITL facility equipped with a single train of a passive safety injection system were performed and their major results were summarized. The effect of the flow distributor, break size and different types of Core Makeup Tank (CMT) and Safety Injection Tank (SIT) on the thermal-hydraulic behavior during a Small-Break Loss of Coolant Accident (SBLOCA) scenario was investigated for the SMART passive safety system development. During the SBLOCA scenario, the water layer was well stratified thermally in the core makeup tank, and the safety injection water was injected efficiently into the reactor pressure vessel from the initial period. Eventually, the reactor coolant system was cooled down properly. From the test results, it was estimated that the SMART passive safety system has sufficient cooling capability to deal with the SBLOCA scenario of SMART. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide the base data during the application phase of the revision of standard design approval and construction licensing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

System Study: High-Pressure Safety Injection 1998–2013

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Improved Safety Injection Flow Map Associated with Target RCS Depressurization to Maintain Core Coolability of OPR1000

In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG, in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the Core Exit Temperature (CET) reaches 923 K. Among 7 mitigation strategies envisioned, the third management strategy, injecting coolant into Reactor Coolant System (RCS) by depressurizing the RCS is one of the key strategies to fortify the in-vessel retention capability. As an aid to such mitigation strategy, a safety injection (SI) flow map has been developed by Lee et al. for predicting required flow rate associated with target RCS depressurization to recover in-vessel coolability. The mechanistic SI flow map used the CET increase rate and core water level decrease rate obtained from MELCOR code simulation. It should be noted that during the severe accident, the total heat generated is sum of the decay heat and the oxidation heat. Thus, it is necessary to consider additional heat sources by oxidation reaction between cladding, structure, control rod, and hot steam prior to the injection of coolant. It was believed that the SI model estimated required flow rate rather conservatively by considering the expected amount decay heat and the accumulated heat in the core based on the RCS pressure. However, it was confirmed that the SI model developed by Lee et al. could not represent the total heat removal especially since start of oxidation. Under transient conditions, the heat sources accumulated in the heat structures (fuel, cladding, supporting structures) should be embarked in the model for accurate contribution of the total heat source to the required flow rate. Therefore, the objective of this study is to improve the previous SI flow map predicting the injection flow rate associated with target RCS depressurization to satisfy the in-vessel coolability during severe accidents. Additional terms were developed to reflect accurate heat accumulating and included in the existing model. Resulting SI flow map includes the target RCS depressurization, which enables discharge of the required flow rate by utilizing the flow rate-pressure curve with operation of two high pressure safety injections (HPSIs). Also recent MELCOR simulation result of a postulated steam generator tube rupture (SGTR) accident verified the accuracy of the improved map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

System Study: High-Pressure Safety Injection 1998-2020

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear reactors. Demand, run hours, and failure data from calendar years 1998–2020 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS), formerly the INPO Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

System Study: High-Pressure Safety Injection 1998-2014

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

System Study: High-Pressure Safety Injection 1998-2022

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 62 U.S. commercial operating nuclear reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hours, and failure data from calendar years 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. Highly statistically significant increasing trends were identified in both the HPSI system start-only unreliability and 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

System Study: High-Pressure Safety Injection 1998-2024

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 62 U.S. commercial operating nuclear reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from calendar years 1998 to 2024 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10-year period and yearly estimates for system unreliability are provided for the entire active period. Statistically significant decreasing trends were identified in both the HPSI system start-only unreliability and 24-hour mission unreliability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

System Study: High-Pressure Safety Injection 1998–2018

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from calendar years 1998 through 2018 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS), formerly the INPO Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Systems Engineering (SE) Approach of Developing A Passive Safety Injection System (PSIS) to Cope with APR1400 Extended Station Black-Out (SBO) Accident Using RELAP5

On 11 March 2011 the power supply and cooling of the three Fukushima Daiichi reactors were disabled by a major earthquake followed by a relatively high tsunami waves. The earthquake damaged the electrical grid which is the normal power supply for the reactors, then the tsunami waves disabled the emergency diesel generators of the Fukushima Daiichi reactors, which left the plant with no alternating current (AC) source. This condition is called 'Station Black-Out'. The reactors were tripped immediately, but due to the nuclear physics phenomena, the reactor cores were still producing thermal power, which is generated by the decay of radioactive fission products in the reactor core, known as decay heat. The decay heat had not been circulated to an external heat sink and eventually the accumulative heat produced in the core led to damage to the reactor cores. After the Fukushima Daiichi accident, Design Extension Condition (DEC) such as total loss of feed water (TLOFW), and prolonged Station Black-Out (SBO) attracted the attention of nuclear experts. The attention focuses on the defense in depth concept, redundancy and sustainability of the systems, and development of the passive safety systems, which rely on natural physics laws. The topic of this paper is to discuss the possibility of attaching a Passive Safety Injection System (PSIS) to the APR1400. The PSIS is a safety system which provides adequate cooling water to the steam generators after the loss of the Auxiliary Feed Water System (AFWS) during an extended SBO accident. The PSIS will be available while the power supply of the plant is being recovered and the active safety systems take over the functionality. The anticipated PSIS will be automatically actuated to enhance the engineered safety feature and to reduce the human errors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermo-Mechanical Analysis of SBLOCA with Loss of One Safety Injection System Using SCDAP/RELAP5

Design Extension Condition (DEC) for nuclear power plants has been being deliberated to bring it within a safety regulation boundary, following the 2011 Fukushima Daiichi accident. Moreover, a new regulation on the LOCA accident analysis designated as 10CFR50.45(c) has been proposed based on a concern of the non-conservatism of current criterion neglecting fuel burn-up effect on the cladding embrittlement. Therefore, a more comprehensive and realistic simulation of fuel rod behavior is required to cope with the stringent environment under LOCA accidents. A SCDAP/RELAP5 system analysis code has been used mainly on the prediction of severe accident phenomena. It has a set of mechanistic models for the fuel deformation, which plays an important role particularly in the early phase progression of the fuel melting. However, limited work has been done focusing on the models only to the fuel failure estimation through the DBA or candidates of DEC. Therefore, some analyses were performed here with a burnup dependent core model to understand the capability of SCDAP/RELAP5 to predict the thermo-mechanical behavior of the fuel cladding such as ballooning, burst and external/internal oxidation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Estimation of Core Recovery Time with TRACE

Following the severe accident in Fukushima Daiichi Nuclear Power Plant in Japan, safety related concerns about the zirconium based claddings used in current light water reactors have increased. High corrosion resistance and low neutron absorption cross section make zirconium a suitable cladding material for normal nuclear plant operational conditions. However, for nuclear accident scenarios when safety functions of the plant are lost, Zr claddings undergo exothermic reaction with steam and exacerbate the accident conditions. The exothermic reaction increases the temperature of the cladding and accelerates core melting in severe accident progression. And the generation of explosive hydrogen gas accumulates and challenges the structural integrity. The accident tolerant fuel (ATF) concept is the fuel and cladding system that gives a longer time margin, compared to standard Zr claddings, to cope with the accident when core cooling systems are lost in a nuclear power plant. The ATF cladding should have slower oxidation kinetics and less hydrogen generation than Zr to alleviate consequences of cladding interaction with steam at high temperatures. The maximum time available to mitigate progression of severe accident is an important measure to evaluate benefits of ATF claddings. For this work, the maximum time available to mitigate progression of severe accident is selected as the time at which cladding melting reaction starts. Melting reaction is selected as an upper limit because of the degradation of coolable geometry. Maintaining the coolable geometry is a regulation criteria for both loss of coolant (LOCA) analyses and outside the LOCA analyses. Once the melting of the cladding starts, it exacerbates the deformation of the fuel assembly and hottest rods within the core cannot be cooled by water reflood and results in fuel melting progression. The aim of this work is to evaluate time margin gain of ATF cladding in a severe accident scenario of LOCA without safety injections and station blackout considering oxidation kinetics only. Also it is assumed that safety injections are recovered before melting of the claddings start and the sensitivity of safety injection actuation time on core recovery is analyzed. The core recovery time is the time when makeup water from safety injections quench the fuel rods and halt the progression of the severe accident. This time is estimated by initiating safety injections at different times and observing the hot rod behavior which represents 1/5. of the core in this work. It is considered that core is not recoverable when 20% of the core is damaged.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2017

Procedure design of F and B operation and its impact on the risk of a nuclear power plant - 168

A feed and bleed (F and B) operation is an operator's emergency task to remove core heat from a reactor coolant system by bleeding coolant from the pressurizer, and feeding makeup water through safety injection systems under the loss of all feedwater condition. Even though the F and B operation is an important task having a significant impact on the risk of a nuclear power plant, there is a high uncertainty regarding its modeling and analysis in a PSA. This paper introduces a study on the design of an operational strategy for an F and B operation and its impact on the risk of a nuclear power plant in Korea. An intensive support analysis was conducted to understand the plant behavior and revise the success criteria for the HRA of the F and B operation. A human performance analysis on the emergency tasks was also conducted using a training simulator to generate the operator's performance time for the F and B operation. We revised the PSA model and updated the risk of the reference plant, which shows that the risk of the revised model is much higher than that of the existing model. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Estimation of LOCA break size using cascaded support vector regression - 334

A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary analysis of core heat-up for ATFs under accident conditions

The accident-tolerant fuel (ATF) system is aim at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the conventional UO{sub 2}-Zr fuel system when the reactor is in normal operation. In this paper, the improved performance of UO{sub 2}-SiC, UO{sub 2}-FeCrAl fuel systems, and of fully ceramic micro-encapsulated (FCM) fuels: FCM-SiC and FCM-FeCrAl, on the core heat-up progress in a pressurized water reactor (PWR) is analyzed by using the best-estimate system code RELAP5 and severe accident analysis program. For this purpose two accident scenarios are selected: a large break LOCA as a design base accident and an extended station blackout (SBO) severe accident. For the LBLOCA with just low-pressure safety injection, larger volumetric heat capacity and thermal conductivity of ATF materials contribute to a larger safety margin in terms of the peak cladding temperature compared with UO{sub 2}-Zr. Thanks to higher melting temperatures, FCM-SiC is failed by melting almost 1200 s later than UO{sub 2}-Zr. The accumulated production of hydrogen for ATF materials is around 10 times smaller than that for UO{sub 2}-Zr. However, high temperature oxidation of SiC cladding generates carbon monoxide which is combustible, while the melting point of FeCrAl is not high enough. In the SBO scenario with mitigation measure, the advantage of ATF is more evident. If the mitigation measure could not be taken, core heatup is inevitable for all types of fuel system, but the temperature rising of ATF core is much slower. In general, improved high-temperature properties and resistance to oxidation play important roles for ATF cores.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗
DOE OSTI2023

Safety Benefits Assessment for Accident Tolerant Fuels in Consideration of Steam Generator Tube Degradation Using Dynamic Event Tree Analysis

Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO 2 -Zr fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗