SEARCH · Engineering Papers

Results for “SHUTDOWN”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 217 records · Page 12

All sources NASA NTRS DOE OSTI

NASA NTRS2006

Flight-Tested Prototype of BEAM Software

Researchers at JPL have completed a software prototype of BEAM (Beacon-based Exception Analysis for Multi-missions) and successfully tested its operation in flight onboard a NASA research aircraft. BEAM (see NASA Tech Briefs, Vol. 26, No. 9; and Vol. 27, No. 3) is an ISHM (Integrated Systems Health Management) technology that automatically analyzes sensor data and classifies system behavior as either nominal or anomalous, and further characterizes anomalies according to strength, duration, and affected signals. BEAM (see figure) can be used to monitor a wide variety of physical systems and sensor types in real time. In this series of tests, BEAM monitored the engines of a Dryden Flight Research Center F-18 aircraft, and performed onboard, unattended analysis of 26 engine sensors from engine startup to shutdown. The BEAM algorithm can detect anomalies based solely on the sensor data, which includes but is not limited to sensor failure, performance degradation, incorrect operation such as unplanned engine shutdown or flameout in this example, and major system faults. BEAM was tested on an F-18 simulator, static engine tests, and 25 individual flights totaling approximately 60 hours of flight time. During these tests, BEAM successfully identified planned anomalies (in-flight shutdowns of one engine) as well as minor unplanned anomalies (e.g., transient oil- and fuel-pressure drops), with no false alarms or suspected false-negative results for the period tested. BEAM also detected previously unknown behavior in the F- 18 compressor section during several flights. This result, confirmed by direct analysis of the raw data, serves as a significant test of BEAM's capability.

Mackey, Ryan↗

NASA NTRS2014

Combustion Stability Characteristics of the Project Morpheus Liquid Oxygen/Liquid Methane Main Engine

The Project Morpheus liquid oxygen (LOX) / liquid methane rocket engines demonstrated acousticcoupled combustion instabilities during sea‐level ground‐based testing at the NASA Johnson Space Center (JSC) and Stennis Space Center (SSC). High‐amplitude, 1T, 1R, 1T1R (and higher order) modes appear to be triggered by injector conditions. The instability occurred during the Morpheus‐specific engine ignition/start sequence, and did demonstrate the capability to propagate into mainstage. However, the instability was never observed to initiate during mainstage, even at low power levels. The Morpheus main engine is a JSC‐designed ~5,000 lbf‐thrust, 4:1 throttling, pressure‐fed cryogenic engine using an impinging element injector design. Two different engine designs, named HD4 and HD5, and two different builds of the HD4 engine all demonstrated similar instability characteristics. Through the analysis of more than 200 hot fire tests on the Morpheus vehicle and SSC test stand, a relationship between ignition stability and injector/chamber pressure was developed. The instability has the distinct characteristic of initiating at high relative injection pressure drop (dP) at low chamber pressure (Pc); i.e., instabilities initiated at high dP/Pc at low Pc during the start sequence. The high dP/Pc during start results during the injector /chamber chill‐in, and is enhanced by hydraulic flip in the injector orifice elements. Because of the fixed mixture ratio of the existing engine design (the main valves share a common actuator), it is not currently possible to determine if LOX or methane injector dP/Pc were individual contributors (i.e., LOX and methane dP/Pc typically trend in the same direction within a given test). The instability demonstrated initiation characteristic of starting at or shortly after methane injector chillin. Colder methane (e.g., sub‐cooled) at the injector inlet prior to engine start was much more likely to result in an instability. A secondary effect of LOX sub‐cooling was also possibly observed; greater LOX sub‐ cooling improved stability. Some tests demonstrated a low‐amplitude 1L‐1T instability prior to LOX injector chill‐in. The Morpheus main engine also demonstrated chug instabilities during some engine shutdown sequences on the flight vehicle and SSC test stand. The chug instability was also infrequently observed during the startup sequence. The chug instabilities predictably initiated at low dP/Pc at low Pc. The chug instabilities were always self‐limiting; startup chug instabilities terminated during throttle‐up and shutdown chug instabilities decayed by shutdown termination.

Melcher, J. C.↗

NASA NTRS2013

Thermal Performance of ATLAS Laser Thermal Control System Demonstration Unit

The second Ice, Cloud, and Land Elevation Satellite mission currently planned by National Aeronautics and Space Administration will measure global ice topography and canopy height using the Advanced Topographic Laser Altimeter System {ATLAS). The ATLAS comprises two lasers; but only one will be used at a time. Each laser will generate between 125 watts and 250 watts of heat, and each laser has its own optimal operating temperature that must be maintained within plus or minus 1 degree Centigrade accuracy by the Laser Thermal Control System (LTCS) consisting of a constant conductance heat pipe (CCHP), a loop heat pipe (LHP) and a radiator. The heat generated by the laser is acquired by the CCHP and transferred to the LHP, which delivers the heat to the radiator for ultimate rejection. The radiator can be exposed to temperatures between minus 71 degrees Centigrade and minus 93 degrees Centigrade. The two lasers can have different operating temperatures varying between plus 15 degrees Centigrade and plus 30 degrees Centigrade, and their operating temperatures are not known while the LTCS is being designed and built. Major challenges of the LTCS include: 1) A single thermal control system must maintain the ATLAS at 15 degrees Centigrade with 250 watts heat load and minus 71 degrees Centigrade radiator sink temperature, and maintain the ATLAS at plus 30 degrees Centigrade with 125 watts heat load and minus 93 degrees Centigrade radiator sink temperature. Furthermore, the LTCS must be qualification tested to maintain the ATLAS between plus 10 degrees Centigrade and plus 35 degrees Centigrade. 2) The LTCS must be shut down to ensure that the ATLAS can be maintained above its lowest desirable temperature of minus 2 degrees Centigrade during the survival mode. No software control algorithm for LTCS can be activated during survival and only thermostats can be used. 3) The radiator must be kept above minus 65 degrees Centigrade to prevent ammonia from freezing using no more than 135 watts of heater power. 4) The LHP reservoir control heater power is limited to 15 watts with a 70 percent duty cycle. 5) The voltage of the power supply can vary between 26 volts direct current and 34 volts direct current during the spacecraft lifetime. A design analysis shows that a single LTCS can satisfy these requirements. However, shutdown of· the LHP is particularly challenging and the shutdown heater must be wired in series with two reservoir thermostats and two CCHP thermostats at different set points. An LTCS demonstration unit has been tested to verify these performance characteristics experimentally prior to proceeding to the final LTCS design and fabrication. Test results showed that the LHP shutdown scheme would be able to shut down the LHP as designed and the reservoir control heater can maintain the ATLAS mass simulator within the plus or minus 1 degrees Centigrade accuracy under various combinations of the heat load, sink temperature, and power supply voltage.

Ku, Jentung↗

NASA NTRSDate not supplied

Ground Test Validations for Uncrewed Dormancy Preparation and Recovery

In future long duration space flight and habitation missions, periods of uncrewed dormancy where life support systems are shutdown or in low power-modes will occur. To examine the impacts of these dormant periods, along with potential pre- and post-dormancy procedures, Marshall Space Flight Center has been building an uncrewed dormancy test stand. The uncrewed dormancy test stand will include a urine separator from the Universal Waste Management System, a condensate separator, urine filter, brine filter and distillate filter. In conjunction to the test stand, the Center for Biofilm Engineering (CBE) has ongoing long-term dormancy experiments to assess microbial survival in a variety of simulated shutdown procedures including shut-in, flushing with and without biocide, and dry-out conditions. This publication will serve as a discussion of dormancy considerations, criteria for success, and experimental designs for testing dormancy conditions and shutdown procedures. It will include an update of ongoing test stand construction, including a description of the current lower-fidelity versions of the urine and condensate separators being produced. It will also examine dormancy work being conducted by the CBE and discuss how this may factor into procedures tested on the uncrewed dormancy test stand.

Eric Beitle↗

NASA NTRSDate not supplied

A Simplified Model of VIPER Thermal Management System. Part I: Loop Heat Pipe

NASA’s Volatiles Investigating Polar Exploration Rover (VIPER) uses a combination of loop heat pipes (LHPs) and heaters as the primary hardware to modulate the temperature of the electronic boxes and payloads instruments. LHPs have been designed and instrumented to be capable of passive and active shutdown. The goal of the passive shutdown, using a thermal control valve (TCV) to limit the flow of the working fluid, is to preserve thermal energy and avoid unnecessary heat leaks during lunar night. The goal of the active shutdown on the other hand, is to forcefully increase the vehicle internal temperature to bakeout the payload instruments. Several numerical modeling tools are available to analyze LHP performance under the complex set of operational requirements and environments levied on the VIPER Thermal Management System (TMS). Yet, while those numerical modeling tools are capable of detailed performance analysis, they tend to take large computational resources and long computational time. In this paper, a model of a single LHP developed based on control volume approach is presented. By using the control volume approach, the complex geometry of the LHP, and the components attached to it, are reduced to thermal lump capacitances, thus reducing the computation effort. This paper also describes the effort to correlate the proposed LHP model using thermal vacuum (TVAC) test data.

Loop Heat Pipe↗

NASA NTRSDate not supplied

A Simplified Model of VIPER Thermal Management System. Part I: Loop Heat Pipe

Loop Heat Pipe↗

DOE OSTI2016

Introduction on Recently Revised Regulatory Framework Related to Decommissioning of Nuclear Facilities in Korea - 16488

There are 24 units of nuclear power reactors in operation and 4 units of nuclear power reactors under construction in Korea as of November 2015. However, there is no permanently shutdown nuclear power reactor and decommissioned or under decommissioning nuclear power reactor. There are only 2 research reactors being decommissioned since 1997. It is realized that improvement of the regulatory framework for decommissioning of nuclear facilities has been emphasized constantly from the point of view of IAEA's safety standards. IAEA published the safety requirement on decommissioning of facilities on July 2014; its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to follow up action on the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, regulatory framework for decommissioning of nuclear facilities in Korea was revised through comparing to IAEA safety standards. It was identified that items should be revised to improve the regulatory framework for decommissioning. Those are as follows: absence of legal definition of decommissioning, incomplete procedure for safety regulation after permanent shutdown, un-detailed acceptance criteria for decommissioning plan, incomplete requirements for early preparing and periodic update of decommissioning plan, un-detailed requirements on standard format and contents for decommissioning plan, and incomplete radiological criteria on site and building reuse after completion of decommissioning. Nuclear Safety Act related to decommissioning of nuclear facilities was revised and promulgated on 21 July 2015. As the lower statute of Nuclear Safety Act, Enforcement Decree of the Nuclear Safety Act and Enforcement Regulation of the Nuclear Safety Act were also revised and promulgated on 21 July 2015. In this paper, related to decommissioning of nuclear facilities such as nuclear power reactor, research or training reactor, and nuclear fuel cycle facility, it was introduced the main changes of the amended and promulgated Nuclear Safety Act on July 2015. It was also mentioned about the current issue in accordance with its implementation. Main contents of revised Nuclear Safety Act are that decommissioning plan should be submitted for nuclear installations to be constructed and operated, and this plan should be updated periodically. In addition, 3 years of grace period was set to submit preliminary decommissioning plan for the facility which has already been approved prior to July 2018. Preliminary decommissioning plan should be updated every 10 years, and regulatory body should review this document. According to the revised Nuclear Safety Act, in the case of a nuclear power reactor after the approval of the change on operating license for permanent shutdown, should submit final decommissioning plan within 5 years, and be approved by the regulatory body. On July 2015, Kori Unit 1 was determined not to apply its 2nd continued operation, and will expire in 2017. Kori Unit 1 will be the first case for the submission of final decommissioning plan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Thermal Model for Horizontal In-Water and In-Air Mini-Plate Experiments in the Advanced Test Reactor

Mini-Plate (MP) irradiation tests are a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). MP experiments are considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. These fueled experiments contain aluminum-clad fuel mini plates consisting of monolithic U-10Mo. Previous thermal analyses have been documented concerning oxide growth[1] and thermal safety margin[2] for the MP-2 experiment. A thermal analysis has been performed on the MP-2 experiment using a new technique for calculating the necessary time after the experiment has been irradiated and reactor shutdown necessary to not exceed the 95% lower bound blister threshold temperature for the fuel in a horizontal in air configuration. A horizontal in water model has also been analyzed and will be discussed. After the experiment is irradiated in the ATR, it is necessary to calculate the fuel meat temperatures at 12 hours after shutdown. This is done to assure that at the 12 hour mark, if this experiment is removed from the reactor and accidentally dropped onto the working tray and is in the most severe geometrical configuration of horizontal (minimum heat transfer) that fission products will not be released. The blister threshold temperature is considered this safety limit. The horizontal in air analysis is also performed considering that the experiment is horizontal in a drained canal and filled with air and natural convection heat transfer with air on the outside. The initial mini-plate testing campaign known as MP-1 [3] has been irradiated and its purpose was to form the base fuel specimens fabricated by various processes so that a single process can be down-selected for qualification. MP-2 will then serve to populate a comprehensive fuel performance data set for the down-selected fabrication process. The MP-2 experiment will be irradiated in several locations of the ATR as shown in Figure 1. The fuel plate thickness and geometry is identical to the MP-1 experiment [3], with the only variable being the fuel meat thickness. This model of the B-12 position is treated as encompassing the I-21, I-24, and B-9 positions since it has more fuel and higher total heat than the other positions. Since the blister threshold temperature is a function of fission density, a series of fission densities is calculated for the B-12 position to cover the other three positions. The B 12 position specifically uses fuel meat that is 0.016 in thick by 3.25 in long and 0.75 in tall. The total plate thickness is 0.050 in. Aluminum cladding forms the barrier between the PCS and the fuel meat. The purpose of this paper is to demonstrate this new technique developed to calculate the necessary decay time after shutdown for a series of fission densities using the ABAQUS [4] finite element and heat transfer code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Activities ONDRAF/NIRAS related to the decommissioning of nuclear facilities

Since 1980, the Agency is responsible by law for the safe management of all radioactive waste produced in Belgium, including decommissioning wastes. By the law of 11 January 1991 and the implementing Royal Decree of 16/10/1991, ONDRAF/NIRAS has been entrusted with a mission concerning the decommissioning of nuclear facilities. This mission involves the collection and assessment of data concerning decommissioning forecasts for nuclear facilities, the approval of facilities' decommissioning programmes, the establishment - in consultation with operators - of financing conditions for decommissioning, as well as the implementation of these programmes on request by the operator, or in the case of its failure to do so. This is the case for the company Best Medical Belgium SA located at Fleurus (MDS Nordion SA, till April 2011), which produced radioisotopes for medical applications and went bankrupt in 2012. These installations have been entrusted to ONDRAF/NIRAS. A plan of action was developed for taking over the operations in the framework of remediation and decommissioning. Steps have been taken to integrate his new role as a nuclear operator. The installations of Best Medical Belgium SA are now referred to as the 'O/N - Site Fleurus.' Nuclear facility operators, or any person requesting to operate a nuclear facility, are obliged to provide ONDRAF/NIRAS, under their responsibility and in due time, with all the necessary information concerning these facilities' decommissioning forecasts, the nature, quantities and dates of transfer of the resulting waste, and the financing conditions for these facilities' decommissioning. In order to make the necessary funds available for decommissioning a nuclear facility when it will be shut down, operators are obliged to establish provisions during the facility's active life. These provisions are calculated in such a way that the total amount established at the time of the final shutdown covers all costs resulting from the facility's final phase, namely the preparation of decommissioning, decontamination, more or less long-term maintenance, dismantling, treatment, conditioning and disposal of waste produced during this phase. In order to face to multiple and repeated evaluation processes, the Agency, already in the early 90's, started with the implementation of its own integrated data processing system, recording the physical and radiological inventories of nuclear facilities, and allowing the evaluation of the quantities of decommissioning materials and wastes as well as of the decommissioning costs of these facilities. The cost evaluations cover all decommissioning activities from final shutdown of the facility until final release from nuclear control, as well as conventional demolition and site restoration if required. Beside the information related to the inventories, the database involves so called auxiliary tables integrating unit prices of the various decontamination and dismantling techniques, unit rates for radioactive waste processing, interim storage and final disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Developing a concept for a national used fuel interim storage facility in the United States

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a 'Monitored Retrievable Storage' facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE's goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Operational Characteristics of an Accelerator Driven Fissile Solution System

Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a “generic” Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system parameters, such as response times, will be quantified. A generalized linear systems analysis of steady-state operations will be performed to evaluate the level of stability of ADFS systems. This information should provide a basic understanding of typical ADFS system operational behavior, and facilitate the development of monitoring procedures and operator aids.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

PWR cores with silicon carbide cladding

The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

ESBWR response to an extended station blackout/loss of all AC power

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackouts for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event. Beyond 7 days, only a few simple actions are needed to cope with the SBO for an indefinite amount of time. The operation of the ICS as the primary success path for mitigation of an SBO, allows for near immediate plant restart once power is restored. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is further surrounded by a substantial 'steel concrete' composite shield building. The containment vessel is not affected by external flooding, and the shield building design provides hazard protection beyond that provided by a comparable reinforced concrete structure. The intent of this paper is to demonstrate the robustness of the AP1000 design against extreme events. The paper will focus on the plants ability to withstand extreme external events such as beyond design basis flooding, seismic events, and malicious aircraft impact. The paper will highlight the robustness of the AP1000 nuclear island design including the protection provided by the unique AP1000 composite shield building. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Application of the JENDL-4.0 nuclear data set for uncertainty analysis of the prototype FBR Monju

This paper deals with uncertainty analysis of the Monju reactor using JENDL-4.0 and the ERANOS code 1. In 2010 the Japan Atomic Energy Agency - JAEA - released the JENDL-4.0 nuclear data set. This new evaluation contains improved values of cross-sections and emphasizes accurate covariance matrices. Also in 2010, JAEA restarted the sodium-cooled fast reactor prototype Monju after about 15 years of shutdown. The long shutdown time resulted in a build-up of {sup 241}Am by natural decay from the initially loaded Pu. As well as improved covariance matrices, JENDL-4.0 is announced to contain improved data for minor actinides 2. The choice of Monju reactor as an application of the new evaluation seems then even more relevant. The uncertainty analysis requires the determination of sensitivity coefficients. The well-established ERANOS code was chosen because of its integrated modules that allow users to perform sensitivity and uncertainty analysis. A JENDL-4.0 cross-sections library is not available for ERANOS. Therefor a cross-sections library had to be made from the original ENDF files for the ECCO cell code (part of ERANOS). For confirmation of the newly made library, calculations of a benchmark core were performed. These calculations used the MZA and MZB benchmarks and showed consistent results with other libraries. Calculations for the Monju reactor were performed using hexagonal 3D geometry and PN transport theory. However, the ERANOS sensitivity modules cannot use the resulting fluxes, as these modules require finite differences based fluxes, obtained from RZ SN-transport or 3D diffusion calculations. The corresponding geometrical models have been made and the results verified with Monju restart experimental data 4. Uncertainty analysis was performed using the RZ model. JENDL-4.0 uncertainty analysis showed a significant reduction of the uncertainty related to the fission cross-section of Pu along with an increase of the uncertainty related to the capture cross-section of {sup 238}U compared with the previous JENDL-3.3 version. Covariance data recently added in JENDL-4.0 for {sup 241}Am appears to have a non-negligible contribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Lessons-Learned from D and D Activities at the Five Gaseous Diffusion Buildings (K-25, K- 27, K-29, K-31 and K-33) East Tennessee Technology Park, Oak Ridge, TN - 13574

The East Tennessee Technology Park (ETTP) is the site of five former gaseous diffusion plant (GDP) process buildings that were used to enrich uranium from 1945 to 1985. The process equipment in the original two buildings (K-25 and K-27) was used for the production of highly enriched uranium (HEU), while that in the three later buildings (K-29, K-31 and K-33) produced low enriched uranium (LEU). Equipment was contaminated primarily with uranium and to a lesser extent technetium (Tc). Decommissioning of the GDP process buildings has presented several unique challenges and produced many lessons-learned. Among these is the importance of good, up-front characterization in developing the best demolition approach. Also, chemical cleaning of process gas equipment and piping (PGE) prior to shutdown should be considered to minimize the amount of hold-up material that must be removed by demolition crews. Another lesson learned is to maintain shutdown buildings in a dry state to minimize structural degradation which can significantly complicate characterization, deactivation and demolition efforts. Perhaps the most important lesson learned is that decommissioning GDP process buildings is first and foremost a waste logistics challenge. Innovative solutions are required to effectively manage the sheer volume of waste generated from decontamination and demolition (D and D) of these enormous facilities. Finally, close coordination with Security is mandatory to effectively manage Special Nuclear Material (SNM) and classified equipment issues. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Potential Cost Implications of an Interim Storage Facility for Commercial SNF+ - 17132

The question of whether or not consolidated interim storage of commercial spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. This paper summarizes an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). In this study, the order-of-magnitude estimates of total system costs were calculated and tabulated. The analyzed scenarios involve shipment of SNF from reactors in dual-purpose storage and transportation canisters (DPCs) currently being used by utilities for dry storage at reactor sites. A number of pertinent conclusions can be drawn from this evaluation, including: - Delay in repository availability increases total system costs. Any delay in opening a repository increases total system costs, regardless of whether the system has an ISF or not. This is due to the increased cost associated with an extended duration for storage, whether at multiple independent spent fuel storage installations (ISFSIs) or at an ISF, until waste can be disposed of.- There is a potential total system lifecycle cost avoidance in realistic scenarios with an ISF when compared to scenarios with no ISF. However, most of the cost avoidance occurs several decades after the ISF is opened. The total WMS cost differential over the long term is mainly attributed to the reduced operational costs of storing the fuel in a consolidated facility rather than at individual reactor sites. - Earlier establishment of an ISF allows for more avoidance of post-shutdown at-reactor storage costs for any repository opening date. An ISF allows earlier acceptance of fuel from reactors, which reduces at-reactor costs from a total system perspective. - Transportation costs have little impact on a WMS with or without an ISF. These impacts range from 3-11% of the total cost in all scenarios. Therefore, transporting the fuel twice does not appear to be a significant cost concern driver relative to other system costs. In conclusion, designing, licensing, and constructing an ISF will require a large near-term financial investment. However, reduced annual operating and maintenance costs for an integrated waste management system with an ISF, compared to a system without an ISF, could eventually result in system-wide (integrated) cost avoidance due to reductions in at-reactor storage costs in the long term. The largest cost avoidances would not occur for several decades, so assumptions about inflation, escalation, and discount rates will have a significant effect on potential economic impacts of an ISF. It should be noted that other benefits associated with the investment of an ISF-such as earlier removal of SNF from shutdown reactor sites, earlier prospects for site re-utilization, and additional overall integrated waste management system flexibility-were not addressed in this study. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗