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DOE OSTI2024

Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Innovative control mechanism for research and test reactors using mandrel-shaped control rods

Research and test reactors have historically played a pivotal role in supporting the initial development of nuclear reactors. They continue to provide essential data for enhancing fuel designs and material knowledge. However, with many such reactors aging and the growing demand for data to bolster advanced reactor development, it is more necessary to research potential design attributes of the next generation of research and test reactors. For test reactors dedicated to fuel and material testing, the design of control mechanisms significantly influences the stabilization of neutron flux levels in irradiation positions while sustaining criticality. This study presents an innovative control mechanism for potential research and test reactor designs. It employs small absorber rods that move in opposite axial directions to maintain axial symmetry of power and neutron flux during burnup cycles. These rods maximize reactivity worth while also offering flexibility to flatten the radial power distribution. An axial translation of the control mechanisms’ absorbers, as compared to the rotational movement of absorbers in control cylinders, also provides a benefit to available excess reactivity and cycle length. Additionally, this work utilizes a simplified core model of the Advanced Test Reactor to assess the performance of this control mechanism. Compared to the current control system based on rotating control cylinders, the new control mechanism has the potential to enhance, or at least maintain, neutronic performance parameters in this reactor design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Assessment of Reactor BN-350 Impact on the Environment During Its Operation Period

Fast neutron reactor BN-350 was designed for dual purposes: steam production and plutonium generation. The steam produced in the reactor was used to generate electricity (150 MW) and desalination of seawater (up to 120 000 tons/day). The BN-350 reactor began operating in July 1973 with a design lifetime of 20 years. After the end of the reactor design lifetime in July 1993, its further operation was managed on the basis of annual decisions taken to extend the operating life for one year. The reactor was in downtime beginning in December 1998 and on 22 April 1999 the RK Government Resolution no. 456 'About decommissioning of the BN-350 reactor in Aktau, Mangistau region' was adopted. Thus, the total actual reactor operational period was more than 25 years. In the framework of work under the approved 'Plan of priority measures for decommissioning of the BN-350 reactor in Aktau', the specialists of the Institute of Nuclear Physics (INP) and the Mangystau Nuclear Power Complex performed a radiation survey of the plant area, the sanitary protection zone and reactor BN-350 monitoring area. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Computational Modeling, Simulation and Analysis of the RPI Walthousen Reactor Critical Facility (RCF)

The Walthousen Reactor Critical Facility (RCF), located in Schenectady, New York, is currently owned and operated by Rensselaer Polytechnic Institute (RPI). The facility was originally constructed by the American Locomotive Company (ALCO) for the research and development of small, portable reactors. Originally designed for highly-enriched uranium fuel, the prototype reactor first came online in August 26, 1956, and was transferred to RPI ownership in 1963. Several changes to the reactor's design and operation have been made, the most notable of which was the conversion from high-enriched uranium to low-enriched uranium fuel. The RCF is an open-pool critical assembly that operates at a low power with a license limit of less than 100 watts. This low operating power permits a much greater flexibility in configuration and operation as compared to other research reactors. Because of this, experiments may be designed that would be unfeasible for research reactors that operate at much higher powers. The RCF's functional capabilities provide opportunities to perform several reactor experiments and measurements, such as: core criticality experiments, differential and integral rod worth measurements, moderator temperature coefficient of reactivity calculations, critical rod height and critical water height measurements, etc.. These experiments are regularly performed in conjunction with RPI course-work education and Senior Reactor Operator (SRO) training. The unique capabilities and experiments that may be performed in the RCF are particularly useful for validating certain aspects of computational analysis codes. A computational model of the RCF reactor has been previously generated and validated using the Monte Carlo N-Particle (MCNP) code. For the purpose of future experiments, this model is also being adapted for use with the SHARP Toolset, which includes the PROTEUS neutronics code, the Nek5000 fluid dynamics code, and the Diablo structural mechanics code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Seismic isolation: A pathway to standardized advanced nuclear reactors

Standardizing advanced nuclear reactors is a pathway to substantially reducing their overnight capital cost and achieving parity with other power sources, including renewables and fossil fuels. The seismic load case has thwarted standardization of nuclear power plants because site-specific seismic hazard and local near-surface geology has triggered soil-structure-interaction analysis, design, equipment qualification, regulatory review, and licensing, ensuring that each build is different. To achieve standardized or site-independent certified advanced reactor designs, the impact of the seismic load case on the engineering and construction cost and time must be substantially mitigated. Seismic isolation is a mature technology that has been used for more than 30 years in non-nuclear sectors to substantially reduce earthquake demands in buildings and other infrastructure. In this paper, seismic isolation is used to enable standardization of advanced reactor designs, aimed at the complete re-use of a site-independent, certified design and repeated procurement of safety-class equipment. A pathway to standardized designs using seismic isolation is demonstrated for two fundamentally different advanced reactors: a molten salt reactor and a high temperature gas reactor. Each reactor building is equipped with three specialized pieces of safety-class equipment, namely, a reactor vessel, a steam generator, and a control rod drive mechanism housing that is attached to the reactor head. Analysis is performed per ASCE and ASME standards to design the buildings and the equipment for two base conditions: conventional (fixed base) and base isolated. The impact of the seismic load case is characterized for the reinforced concrete walls in the buildings and for the equipment, measured using vessel wall thickness and horizontal accelerations. Here the analysis results show that the fixed-base buildings, designed for a site of low seismic hazard (peak ground acceleration, PGA = 0.15 g) could be constructed at a site of much greater seismic hazard (PGA = 0.7 g) if seismic base isolation is employed. Importantly, the scope of the site-specific analysis, design, and qualification would be limited to the seismic isolators and the isolated substructure, drastically reducing plant-specific engineering, review, and licensing, and time to construction start. Regulatory challenges and opportunities with standardized reactor designs are identified.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

A brief history of design studies on innovative nuclear reactors

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Accident analysis of heavy water cooled thorium breeder reactor

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Temperature and neutron flux dependence of in-reactor creep for cold-worked Zr-2.5Nb - 2016-0042

Over the past 10 years, internally pressurized capsules made from Zr-2.5Nb tubing have been irradiated in the National Research Universal (NRU) reactor at Chalk River Laboratories at temperatures of 280, 320, and 340 deg. C and dose rates between 3 X10{sup 15} n. m{sup -2}. s{sup -1} and 2 X10{sup 17} n. m{sup -2}. s{sup -1} (E>1 MeV). Periodic gaging has been used to assess the primary and secondary (steady-state) creep behavior. The objective of this detailed and controlled experiment was to determine, for the first time, the creep and microstructure evolution in Zr-2.5Nb tubing over a wide range of irradiation conditions for fast neutron fluxes applicable to a CANDU pressure tube. Similar but 'accelerated' creep experiments have been conducted in the Osiris test reactor at fast neutron fluxes of approximately 1.8 X10{sup 18} n. m{sup -2}. s{sup -1} (E>1 MeV), much greater than the neutron fluxes in the NRU reactor. Although accelerated tests in high-flux reactors such as Osiris provide information on irradiation creep, they do not represent the neutron flux conditions applicable to a power reactor. Tests covering power reactor operating conditions are needed to develop models for in-reactor creep of pressure tubes under the appropriate conditions. The data from the NRU reactor are compared with results from creep capsules with similar starting microstructures but irradiated in the Osiris reactor. The results show that the steady-state diametral and axial creep rates have a complex dependence on stress, temperature, and fast neutron flux. Data from out-reactor creep tests on unirradiated and pre-irradiated creep capsules that show the effect of prior irradiation on creep are also reported. The results are discussed in terms of a combination of creep mechanisms involving dislocation glide and mass transport. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Assessment of Nuclear Sensors and Instrumentation Maturity in Advanced Nuclear Reactors

In the last decade, 97% of the worldwide commercial nuclear reactors connected to the grid were Light Water Reactors (LWRs). LWRs are expected to stay the dominant type of nuclear reactors for the next few decades. Reliable and redundant safety systems are required in nuclear reactors to ensure safe operation and shutdown in abnormal conditions. These safety systems are actuated by the signals obtained from several sensors and instrumentation in and out of the reactor core. Research and Development (R&D) in advanced sensors and instrumentation has gained extra attention, particularly following the accident at the Three Mile Island Unit-2 (TMI-2). In LWRs, these sensors and instrumentation have shown a high level of maturity with long operating experience. Ensuring the compatibility of these sensors and instrumentation with advanced nuclear reactors (Generation IV) is necessary, particularly with the expected expansion of the nuclear industry in the next few decades. Nuclear Sensor and instrumentation technologies used in the current generation of LWRs were investigated. The compatibility of these technologies with advanced reactors was assessed by comparing the advanced reactors' environments with those of the currently operating reactors. In addition to that, the needed R&D for such technologies was highlighted. In comparison with the LWRs environment, it was shown that advanced reactor environments are expected to experience elevated temperatures, a fast neutron spectrum, and a harsh corrosion environment. It was demonstrated that R&D is required mainly for fixed in-core nuclear sensors and instrumentation, while it is not a priority for ex-core nuclear sensors and instrumentation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Remote-Controlled Removal of Bricks from a Reactor Shaft: An Unconventional Approach - 16406

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. The complete dismantling is based on 10 decommissioning permits. Package 3 of the 9. decommissioning permit, which is currently under preparation, comprises the dismantling of the biological shield. Package 2 comprised the dismantling of the Primary Shield. In preparation to lift the 14 primary shield segments on height of the reactor core, these were milled at 16 positions and equipped with a total of 26 holes. As a result of the milling and drilling process, an estimated amount of 500-600 kg cast iron swarfs accumulated at the bottom of the 12 m deep reactor shaft. The cobalt in the cast iron was activated to cobalt-60 during operation. This results in a measurable dose-rate of 30-35 mSv/h. At the conical bottom of the reactor shaft were still the remaining parts of the thermal insulation consisting of a lining with brick stones, covered with loose remnants of the thermal insulation, loose metallic residues and cut-off wheels of the angel grinder. For the next dismantling step, the removal of the activated part of the biological shield, it is necessary that the cast iron swarfs must be removed as well as the remains of the thermal insulation. Firstly, this is necessary to reduce the dose rate inside the shaft for subsequent breakthrough operations and, secondly to have a statically resilient surface for the installation of a necessary working platform for next dismantling tools. There was no access for personnel. A manual removal was also not possible due to the existing dose rate. The cast iron swarfs supposed to be removed by a remote controlled suction unit, which was installed on top of a drum and mounted in a special rack. The complete system was lowered into the reactor shaft by using a crane. After removing the major amount of swarfs, the remaining swarfs as well as the sheet metal liner should be collected with a magnetic grabber. The solution for the metal removal was easy to realize, but there was no idea how to remove the remaining bricks remote controlled out of the reactor shaft. Another boundary was that only existing facilities were supposed to be used in addition to the new suction system, i.e. the existing cranes, gripper, as well as the double lid lock of the encasing for the handling of the vacuum cleaner, the transport of the drum and the lock procedure. The design essentially consists of a suction unit and a collecting unit. Both units are connected hooked by the three single suspension chains on the rack on the traverse and connection of the suction tube. Subsequently the entire unit was drained in the 12 m deep reactor shaft. The entire unit has been lifted out of the reactor shaft and placed on a defined set-down position. At this position, now both of the units could be separated by using the Master-Slave-Manipulator in the encasing (A100). By using different suction pipes the reactor shaft bottom could be exempted from the cast iron swarfs within a few days. The suction unit worked in such a positive way that it was decided to remove the loose refractory bricks, metallic residues, cut off wheels, etc., too. Metal parts remaining on the ground were removed by a magnet. This pragmatic approach avoided additional complex design works and the need of constructing special solutions and thus, massively saved time, cost and dose rate. The duration of planning, permission including implementation amounted to approximately 3 to 4 months. The additional work could be performed directly after the actual dismantling step.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Shielding Design for Prototype Generation-IV Sodium-Cooled Fast Reactor

A Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) with 150 MWe power rating is being developed jointly by the Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL). KAERI is responsible for the overall project with the ANL contributions integrated into the KAERI activities. Shielding concept of PGSFR was developed to ensure that the safety limit on internal structural damage and working personnel received dose is not exceeded. The first part evaluated the Displacement Per Atoms (DPA) and fast neutron fluence on the structures inside the reactor vessel, such as lower grid, Upper Internal Structure (UIS), barrel, Redan, and reactor vessels. The analysis determined the thickness of shields required to satisfy the DPA and fast fluence limit. The second part of the shielding analysis evaluated the radiation dose rate above the reactor head due to the primary sodium activation inside the reactor vessel. The analysis determined the necessary thickness of reactor head and additional shield to ensure that the dose rate received by personnel working on top of the reactor is below the safety limit. The last part of the shielding analysis investigated the dose rate received by personnel working near the secondary loop. The secondary sodium is activated inside the Intermediate heat exchanger (IHX). The activated secondary sodium flows out of the reactor vessel and emits high energy photons along the way. The shielding designs both inside the reactor vessel and around the pipes of the secondary loop were investigated. The shielding concept for the PGSFR was developed to minimize radiation exposure to plant personnel and the irradiation damage of the internal structural components. The damage of the internal structural components are mainly due to the loss of the ductility by the neutron irradiation, and the activated gammas of primary and secondary sodium coolants are the major radiation sources that are required to be shielded. Thicker and denser shielding structures are favorable to reduce the radiation dose and protect personnel who work in the plant; however, which enlarges the in-vessel space to accommodate the shielding structures. Thus, in order to install the shielding structures in a reasonable reactor to satisfy the DPA or dose rate limit size, the PGSFR shielding parameters were determined from the sensitivity study, which includes the compositions of the shielding materials and dimensions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Effect of Reactor Radiation on the Thermal Conductivity of Graphite-based Dispersion Fuel

The Fukushima Daiichi nuclear accident in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is highly preferable to simulate off-normal transients in addition to standard in-core fuel performance tests (available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory). The Transient Reactor Test Facility (TREAT) is now being prepared to restart its operation, performing transient testing to assist advanced fuel development. TREAT was operated from 1959 to 1994, and was extensively used to generate thousands of reactor transients for different programs. The reactor has not operated since 1994. The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium). For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO{sub 2} contains only low-enriched uranium (LEU). The preliminary LEU core design has shown a much lower carbon/uranium atomic ratio and enrichment than in the HEU core design. To achieve performance similar to the HEU core, the prospective LEU core will contain a much higher density of UO{sub 2} particles. As a result, a higher level of radiation damage, particularly from fission fragments emerging from the UO{sub 2} particles, can be expected dependent on the UO{sub 2} particle size within the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled, and validated based on the available experimental data. In this study, the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed and validated. The proposed method can be further applied to predict the changes in the thermal conductivity of the prospective LEU TREAT fuel during long-term operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Adjusted Graphite Thermal Scattering Data Applied to Fluoride-Salt-Cooled Reactors

Thermal scattering data is a key component of fission reactor simulation. While most materials can be reasonably modeled with a free gas approximation, the moderators in thermal systems require a more rigorous treatment to sufficiently capture their specific neutron interaction physics. Graphite does not typically feature as a principle moderator in power reactors in service today, but remains an attractive prospect for a number of new reactor concepts currently in development. In comparison to light water, a material nearly ubiquitous in thermal reactors, graphite offers a better neutron economy by way of a small absorption cross section, high thermodynamic stability and can be incorporated into the structure of the core itself. Graphite in its purest form is a crystalline material composed of stacked layers of graphene, a hexagonal planar lattice of carbon atoms. In-plane, each carbon atom is covalently bonded to 3 neighbouring carbon atoms, creating a strong bond, while between planes only a weak van der Waals force exists. Pure graphite has a density of 2.26 g/cm{sup 3}. However, reactor-grade contains a significant amount of carbon binder, leading to a porous bulk medium with randomly oriented micro crystals. Densities of 1.5 - 1.9 g/cm{sup 3} are typical for reactor-grade graphite, corresponding to porosities on the order of 10-30%. This deviation from pure crystalline graphite should be accounted for in the thermal scattering data. When an incident neutron is in or near the thermal region, its de Broglie wavelength is the same order as the molecular bond lengths, leading to elastic scattering. At similar energies, the dynamics of the moderator lattice (in the case of solid moderators) can cause inelastic scattering. For these moderators, special cross sections should be calculated and used in reactor simulations. To generate this thermal scattering data, parameters pertaining to the properties and structure of the material in question are fed into a nuclear data processing code, such as NJOY. Within NJOY is a module called LEAPR which is used to generate the thermal scattering law S(α,β), which is in turn is fed into another module THERMR which is used to generate thermal cross sections that factor in the molecular-scale scattering phenomena. The core of the thermal scattering law is the phonon spectrum ρ(β) which describes the modes of excitation of the material lattice. An investigation into the phonon frequency spectrum was conducted at North Carolina State University (NCSU) for reactor-grade graphite which accounts for porosity. This phonon spectrum was calculated using the ab initio density functional theory (DFT) code VASP. The inelastic cross section calculated using this spectrum was found to have good agreement with relevant experimental data. The resulting spectra differed from the benchmark spectrum produced by Young and Koppel, the same data used in current nuclear data evaluations. The latter uses the lattice dynamics of perfect crystalline graphite with four force constants fitted to the model from thermodynamics data from reactor-grade (porous) graphite, which yields a reasonable approximation to the phonon spectrum of reactor-grade graphite.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O 2 fueled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Westinghouse Small Modular Reactor nuclear steam supply system design

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of Remote Dry Cutting Technology for Reactor Vessel Segmentation - 19176

Decommissioning of nuclear power plants is becoming a large market due to the many reactors reaching their operational lifetime limit in the near future. One of the first activities in a decommissioning project is to segment the reactor internals and the reactor vessel. These activities are also one of the more complex ones since they deal with most of the radioactivity inventory that is left in the plant after the spent fuel is gone. They are almost always on the critical path, so an improvement in execution time will be favorable for the entire decommissioning program. Westinghouse has been successful for more than 30 years in more than 40 segmentation projects globally. The traditional technology that has been used in these projects has up until now been underwater mechanical cutting of both the reactor internals and the reactor vessel. Underwater mechanical cutting is proven and very reliable, but it is slow and since there are major advantages to the decommissioning program with improvements in schedule performance, Westinghouse is developing alternative technologies both as part of current offerings but also as entire new concepts. For reactor vessel cutting, Westinghouse has developed a new concept involving thermal cutting in a dry environment. The benefit of dry cutting is the speed of the cutting and the subsequent schedule and cost improvements. A dry environment also means that the pools can be drained which means that a critical activity can be moved forward in the schedule and that there will need to be less monitoring and controlling with no pumps, cooling and recirculation system that need to be operated. Going from underwater mechanical to dry thermal cutting reduces the on-site schedule by more than 50%. The dry cutting concept development considers both PWR and BWR reactors. Some of the challenges of dry thermal cutting involve the creation of radioactive off-gases and how to ventilate these, the extreme heat generation, remote handling of the thermal equipment, cutting station arrangements and how to remove the insulation around the reactor vessel. Depending on the concept the reactor vessel might need to be lifted from the pit in one piece, which also poses some challenges due to the weight, volume and radioactivity of the vessel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗