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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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18 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

System Study: Residual Heat Removal 1998-2014

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 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 trends were identified in the RHR results. A highly statistically significant decreasing trend was observed for the RHR injection mode start-only unreliability. Statistically significant decreasing trends were observed for RHR shutdown cooling mode start-only unreliability and RHR shutdown cooling model 24-hour unreliability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

System Study: Residual Heat Removal 1998-2020

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and in response to post-trip shutdown cooling) at 104 U.S. commercial nuclear reactors. Demand, run hour, and failure data from 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 RHR results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

System Study: Residual Heat Removal 1998-2022

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and in response to post-trip shutdown cooling) at 93 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 1998–2022 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. 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 RHR results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

System Study: Residual Heat Removal 1998–2018

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 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 RHR results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2026

System Study: Residual Heat Removal 1998-2024

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and in response to post-trip shutdown cooling) at 93 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 1998 to 2024 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. 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 and highly statistically significant decreasing trends were observed for RHR low-pressure injection mode start-only unreliability and RHR low-pressure injection model 24-hour unreliability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

System Study: Residual Heat Removal 1998–2013

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 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 trends were identified in the RHR results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗
DOE OSTI2015

Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Preliminary Design of Critical Function Monitoring System of PGSFR

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation control and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Research and Experiment on the Key Technologies of the Spent Fuel Storage in HTR-PM - 18486

High Temperature Reactor demonstration project of China (HTR-PM) is one of the major national science and technology projects, some key technologies for the design and operation of the demonstration project are formed to research according to the main technology. The spent fuel storage system is one of the main systems of the HTR-PM. In order to solve the technical problems related to the operation of spent fuel, several experiments have been conducted, including: spent fuel loading, getting back, spent fuel canister hoisting, storage well cover hoisting, storage canister drop and so on. In the study process, the test facility has been set up to carry on the operation and the experiment verification to ensure the reliability and safety of the spent fuel storage system. Spent fuel storage system undertakes spent fuel receiving, spent fuel canister transfer, storage, lifting and other functions in HTR-PM, the project research need to solve several major technical problems related to the spent fuel operation, which mainly include: Develop the transportable spent fuel canister for HTR-PM, the canister is corrosion resistance, impact resistance, and is in subcritical condition when loaded with fuel. Develop the spent fuel canister hoister, which could hoist the spent fuel canister safely and reliably. Develop the spent fuel storage well cover hoister, which could hoist the well cover safely and reliably. Study the small probability of the spent fuel canister in extreme condition of canister hoist falling accidents, to ensure the radiation protect safety and residual heat removal safety of the canister in case of the canister drop when lifting. Develop the spent fuel charging device and return device, which could load the spent fuel elements into the spent fuel canister, and could take out the fuel elements from the canister. The key technology and equipment of the spent fuel storage system in HTR-PM would be solved through the research of the project. Project research is very important for the implementation and safe operation of the HTR-PM project. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Information on the Advanced Plant Experiment (APEX) Test Facility

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, "Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing," Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Performance Analysis of the Passive Residual Heat Removal System of an Integral Reactor

A demand for small modular reactors (SMRs) is gradually emerging worldwide, especially in Middle East Asia and Africa. SMART (System-integrated Modular Advanced ReacTor) developed by KAERI (Korea Atomic Energy Research Institute) is the first mover among several SMRs. It is designed to produce 100 MW of electricity, or 90 MW of electricity and 40,000 tons of desalinated water concurrently, which is sufficient for 100,000 residents. SMART is currently holding the initiative as the first licensed integral reactor in the world since 4 July 2012, when the standard design approval (SDA) was officially issued by the Nuclear safety and security commission (NSSC). After that, the SMART safety enhancement project was successfully conducted with intense support of the Government, resulting in an in-depth development of the most advanced safety features including the passive residual heat removal system (PRHRS). Recently, an agreement between KAERI and King Abdullah city for atomic and renewable energy (K.A. CARE) was officially signed for the pre-project engineering (PPE) in pursuit of the construction of SMART and developing human resources capability. Both parties are currently conducting a three-year preliminary study to review the feasibility of building SMART and to prepare for its commercialization. KAERI also takes responsibility to train R and D personnel chosen from Kingdom of Saudi Arabia. Sooner or later, SMART will open the global market of SMR and be the best competitive pioneer. Since the Fukushima accident, the passive cooling system of a nuclear reactor has received a great amount of attention. SMART is designed with passive systems in order to ensure the safety of the reactor. The PRHRS is one of the passive safety systems, which is activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators (SGs) until the condition of the RCS reaches a safe shutdown condition. In the present study, the thermal performance of the PRHRS is predicted by using in-house code. Temperature of the RCS and the cooling rate of the RCS are traced with time to see if it is satisfactorily designed to meet the design requirements of the PRHRS. The dependency of the mass flow rate of the natural circulation on the filling ratio (FR) is also discussed. The FR is arbitrarily chosen from 0.22 to 0.50 and further parametric study is needed in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of safety requirements for HTGRs design - 18467

The safety requirements for the design of High Temperature Gas-cooled Reactors (HTGRs), which is analogous to the IAEA specific safety requirements (SSR) for Light Water Reactor (LWR) design (IAEA SSR-2/1), has been developed by the research committee established in the Atomic Energy Society of Japan so as to incorporate the HTGR safety features demonstrated by High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the coupling of the hydrogen production plants with nuclear plant. The safety design approach was determined to establish a high level of safety design standards by utilizing inherent safety features of HTGRs. The safety requirements is then developed using the objectives provisions tree methodology based on the safety design approach. The requirements for a structural integrity of fuel during accident conditions, for a confinement structure instead of a leak-tight pressure resistance containment vessel, for a passive residual heat removal from outside of reactor pressure vessel, etc., which are significantly different from those of LWR, are stated in the developed safety requirements. This paper describes overview of the proposed HTGR specific safety requirements. (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 OSTI2018

Characteristic Atmospheric Radiative Heating Rate Profiles in Arctic Clouds as Observed at Barrow, Alaska

Abstract A 2-yr cloud microphysical property dataset derived from ground-based remote sensors at the Atmospheric Radiation Measurement site near Barrow, Alaska, was used as input into a radiative transfer model to compute radiative heating rate (RHR) profiles in the atmosphere. Both the longwave (LW; 5–100 μm) and shortwave (SW; 0.2–5 μm) RHR profiles show significant month-to-month variability because of seasonal dependence in the vertical profiles of cloud liquid and ice water contents, with additional contributions from the seasonal dependencies of solar zenith angle, water vapor amount, and temperature. The LW and SW RHR profiles were binned to provide characteristic profiles as a function of cloud type and liquid water path (LWP). Single-layer liquid-only clouds are shown to have larger (10–30 K day−1) LW radiative cooling rates at the top of the cloud layer than single-layer mixed-phase clouds; this is due primarily to differences in the vertical distribution of liquid water between the two classes. However, differences in SW RHR profiles at the top of these two classes of clouds are less than 3 K day−1. The absolute value of the RHR in single-layer ice-only clouds is an order of magnitude smaller than in liquid-bearing clouds. Furthermore, for double-layer cloud systems, the phase and condensed water path of the upper cloud strongly modulate the radiative cooling both at the top and within the lower-level cloud. While sensitivity to cloud overlap and phase has been shown previously, the characteristic RHR profiles are markedly different between the different cloud classifications.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Flow Characteristics of Intermediate Heat Exchanger with CFD Methodology on Nuclear Safety Application

The intermediate heat exchanger (IHX) plays an important role in the field of nuclear power plant, and its safety margin could strongly affected by the flow distribution of secondary flow inlet. As a result, the size of secondary inlet is crucial in thermal design. However, the IHX contains thousands of heat exchanger tubes, and it takes larger time to simulate each tube accurately. Other than that, there are more delicate devices in a nuclear power plant, so it would take too much time to simulate actual model. Although system code could solve problems effectively, it could not detect local deformation. On nuclear safety application, how to save time and receive specific results in an efficient way is necessary. Therefore, the porous media model and the heat exchanger model from computational fluid dynamics (CFD) were adopted in this study. The methodology not only can save much time than actual model but also receive more specific results than system code. To ensure the accuracy, validation of the methodology is also presented in this study by modeling RHR heat exchanger of L nuclear power plant. The results of the simulation were compared to experimental values from L nuclear power plant and receive small errors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Robotic Underwater Decontamination Services for Nuclear Applications-18464

Successful Underwater Robotic Decontamination Demonstrates Dose-Saving And Outage-Shortening Technology Diakont successfully provided underwater robotic decontamination services on a North American nuclear power plant's refueling cavity and dryer-separator pool in Spring 2017 using a new tool system. Historically, nuclear plant operators have conducted cleaning and decontamination of these surfaces manually, after draining water from the space. However manual decontamination is slow, and can result in excessive personnel dose exposure. The underwater robotic decontamination services present a vast improvement over manual decontamination because it reduces personnel dose exposure, reduces radwaste, doesn't impact plant chemistry, and doesn't risk inadvertently spreading contamination. Also the new decontamination method avoids the risk of personnel injury and component damage associated with hydrolasing. And by performing the decontamination robotically while the cavities are flooded, in many cases the critical path outage schedule can be shortened. Able to decontaminate horizontal, vertical, and curved surfaces, the decontamination tools easily navigate to areas within flooded cavities that are inaccessible to previous solutions, performing decontamination in parallel to other activities including fuel movement. Also, unlike other legacy solutions, the new tools do not require continuous use of an overhead crane or other method of suspension while performing the decontamination activities. The ROV-type decontamination tool attaches and drives along the cavity and component surfaces using a high-force, no-flow vortex generator, even in the presence of Residual Heat Removal (RHR) or shutdown cooling flow. Efficient, effective cleaning is performed using a rugged brushing action to detach the crud, while vacuuming it away at high flow rates to a submerged filter. At the Spring 2017 refueling outage, the decontamination services with the new methodology were so effective that no additional manual cavity decontamination was required after drain-down. Using the tool's ability to swim, attach, and crawl, field technicians decontaminated the majority of the surfaces designated by the plant operator. Preliminary surveys indicated that all contamination levels were reduced to <50 K dpm/100 cm{sup 2}, helping the utility meet their INPO/Industry collective radiation exposure goals. The new cleaning and decontamination services are ideal for servicing various underwater areas within nuclear power plants such as refueling cavities, BWR-6 drywell heads, spent fuel pools, cask loading pools and fuel transfer canals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗