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116 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Measuring, Cutting and Sorting Irradiated Control Rods to Limit the Volume of Intermediate Level Long-lived Waste - 17001

During operation, the PWR control rods are either completely extracted (scram or safety rods) or partially inserted in the operating zone of the core (fine control or operating rods). At the nominal rating conditions or close to them, the operating zone is limited to some steps of insertion in the order of a few tens of cm. In fact, only the lower part of the control rods are subjected to a significant neutron flux with an impact on the neutron activation and so on the waste classification (Low Level Short-lived Waste versus Intermediate Level Long-lived Waste). Our calculative as well as experimental experience confirms the significant presence of radionuclides in the control rods, but only in the lower parts. Measuring, cutting and finally sorting control rod scraps can reduce the volume of Intermediate Level Long-lived Waste planned for our storage deposit facility. This paper presents the EDF/DP2D methodologies followed to determine the sort criteria between the Low Level Short-lived Waste versus the Intermediate Level Long-lived Waste zones for the Silver-Indium-Cadmium control rods. The irradiated control rods are stored in cases without their head. Each case can contain several control rods, and the cases are stored under water in the fuel assembly pool. To calculate the radioactive inventories, we have developed a calculation scheme which is composed of a 3-dimensional mapping of the neutron flux, a calculation of the evolution of the isotopic compositions under neutron flux and finally a comparison between calculated and measured values. The main hypotheses of the control rods calculation take into account: - 3-dimensional geometry and 3-dimensional neutron sources distribution, - the isotopic compositions including impurities, - the history of irradiation resulting mainly from the daily power production, the positions of the control rods in the loading patterns and the positions in the axial operating zone. In order to validate the calculations, measurements were made 20 meters under water in the fuel assembly pool. The experimental results show the presence of many gamma emitter radionuclides. Only Ag-108m was retained for validating the calculations. This is due to the fact that silver is considered as the 'standard' chemical element for our control rods and silver produces Ag-108m which is a long-lived gamma emitter radionuclide. The analysis of the calculated results and the measured values allow us to build a simplified physics model for the simulation of the control rods activation. Finally three sort criteria can be proposed: - Without measurement devices, the sorting is based on an empirical sort criterion. Cuttings must then be made on the 2 back-ends of the control rod pins. This option is due to the difficulty in knowing the orientation of the pins in the storage cases. The cutting lines are placed at approximately 1 m from the 2 back-ends of the pins. This approach does not allow a great optimization of the sorting, less than 50% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. - With a simple dose rate measurement device, the knowledge of the orientation of the pins in the storage cases allows us to limit cuttings to the lower part of the control rods. Depending on the 'top to tail' cases, between approximately 50% and 75% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. - With an accurate measurement device, the sorting can be based on a specific activity of Ag-108m. Over 4400 Bq/g of Ag-108m (corresponding to the Intermediate Level Long-lived Waste limit), the control rod pieces can be considered as a long-lived waste. The measurements must be made over the 2 back-ends of control rod pins to examine the cases of 'top to tail' storage. This approach allows a great optimization of the sorting. Depending on the history of irradiation, more than 75% of the length of the control rods can be saved from Intermediate Level Long-lived Waste storage. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Potential Impact of Cladding Wettability on LWR Transient Progression

The US Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance. This type of fuel is also known as accident tolerant fuel (ATF). Two promising cladding material candidates with potentially enhanced accident tolerance are silicon carbide (SiC) and an iron-chromium-aluminum (FeCrAl) alloy. The FeCrAl alloy features significantly enhanced oxidation resistance versus the Zr-based alloys currently used in state-of-the-art light water reactors (LWRs). Thermal-fluid characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One characteristic related to boiling heat transfer is the critical heat flux (CHF). At CHF, the efficacy of heat transfer rapidly degrades and components can heat up rapidly. CHF is the heat flux at which the departure from nucleate boiling (DNB) occurs in PWRs. DNB is directly relevant to the departure from nuclear boiling ratio (DNBR), which is the ratio between the critical heat flux and the actual heat flux. Surface wettability is an important factor that influences CHF. CHF plays a significant role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. Significant changes in CHF properties and boiling heat transfer characteristics will impact the progression of transient and accident events and reactor safety characteristics. Additionally, the behavior of the candidate cladding materials post-CHF impacts reactor safety. A brief description of the relationship between cladding material properties and CHF is provided, followed by an analysis of two hypothetical transient scenarios: a super-prompt reactivity-initiated accident (RIA) and an anticipated transient without scram with instability (ATWS-I). The analysis illustrates the potential impact of cladding material properties on safety margins via direct alteration of calculated CHF values. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Modeling PWR systems with a unique framework for hybrid energy systems

Hybrid energy system (HES) is used to utilize the wind, solar, nuclear and other energy sources. Within the study and design of a nuclear power plant and HES, extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including unanticipated transients without scram and hypothetical accidents. A modelling tool that is capable of solving governing equations of mass, momentum and energy for two-phase application is necessary to model the coolant systems of HES. Therefore, this study shows the development of a new coupling tool that is capable of modelling several components of HES and experimental validation of the model. For the initial development phase, the primary and secondary coolant systems (which use several components similar to the components of HES) are modelled by using this new tool. Code-to-code benchmark is also used for the performance evaluation of the new tool. It is observed that there is a good agreement between the new tool and RELAP5 predictions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

ViBRANT Applied as a Closed Loop Flux and Thermal Surrogate

Surrogate testing platforms provide a means to develop a target aspect of a larger system. Some systems target single aspects while others are more wholistic. ViBRANT (Visual Benign Reactor as Analog for Nuclear Testing) aims to be a wholistic system by representing the primary reactor behaviors through an accessible LED driven reactor core, photodiodes, thermocouples, and high-performance software interfaces being developed to the standards required for actual microreactor deployment such as MARVEL. With the ability to drive control drums enabled with full reactor capability (e.g. speed/angle/torque capability/limits for operation and clutch/spring scram capability), drive ViBRANT’s high fidelity 36 pin hexagon core to represent phenomena via LEDs, and capture the analog response of distributed photodiodes/thermocouples, this paper will show the system’s ability to operate in a closed-loop fashion and do well at matching reactor power/temperature distribution in an accessible reactor surrogate environment.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

NEA HTTR LOFC Project Test#2 Benchmark Results

The High Temperature Engineering Test Reactor (HTTR) is a 30 MW prismatic high-temperature gas-cooled reactor (HTGR) owned and operated by the Japan Atomic Energy Agency (JAEA). Staring in 2010, HTTR was used in a series of three loss of forced cooling (LOFC) tests without SCRAM to demonstrate the inherent safety of HTGRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 2, April-June 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 97 The Nuclear Community and the Public: Cognitive and Cultural Influences on Thinking About Nuclear Risk, M. A. Meyer; 109 Twenty-Third Water Reactor Safety Information Meeting, D. A. Copinger; ACCIDENT ANALYSIS: 126 Analysis of a PWR LBLOCA Without SCRAM, Trevor N. Tyler, Rafael Macian-Juan and John H. Mahaffy; DESIGN FEATURES: 139 Vulnerability of Multiple-Barrier Systems, N. C. Lind; ENVIRONMENTAL EFFECTS: 149 A Study of Wet Catalytic Oxidation of Radioactive Spent Ion Exchange Resin by Hydrogen Peroxide, Xingchao Jian, Tianbao Wu, and Guichun Yun; 157 A Comparison Study and Resolution of Differences Between Emergency Response and Safety Analysis Codes Used at the Savannah River Site, A. A. Simpkins; OPERATING EXPERIENCES: 164 Reactor Shutdown Experience, Compiled by J. W. Cletcher; RECENT DEVELOPMENTS: 167 Reports, Standards, and Safety Guides, D. S. Queener; 172 Proposed Rule Changes as of Dec. 31,1995; ANNOUNCEMENTS: 178 American Nuclear Society 1997 Annual Meeting; 178 American Nuclear Society Nuclear Criticality and Safety Division Topical Meeting; 176 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Concept for LEU Burst Reactor

Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗