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DOE OSTI2022

Controlling a power output of a nuclear reaction without control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗

DOE OSTI2024

Controlling a power output of a nuclear reactor without using control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above there actor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗

DOE OSTI2019

Preliminary In-Vessel Shielding Analysis for the Versatile Test Reactor

Shielding design is an important part of the Versatile Test Reactor (VTR) design. Various shielding implementations are required to protect workers and reactor components from damaging radiation sources. The major radiation source in VTR during operation is the prompt neutron and photon emitted from nuclear fuel fissions as well as the activated structure and coolant. Shields are needed to protect the structures located inside the reactor vessel, such as the lower grid, the barrel wall, etc., in order to limit the degradation of the materials properties as a result of neutron damage (“displacement per atom”). Another early key challenge of shielding design for a pool-typed Sodium-cooled Fast Reactor (SFR) such as VTR is the activation of the secondary sodium as to flows through the intermediate heat exchangers (IHX). For a pool-typed SFR, the IHXs are placed in the primary sodium pool inside the reactor vessel. During operation, neutrons produced in the core can reach the IHXs due to the small absorption cross section of sodium and the large mean free path of fast neutrons. The secondary sodium gets activated inside the IHX and carries some of that activity outside of the reactor vessel. This source of radioactivity contained in the secondary sodium needs to be mitigated to meet the radiation dose limit sets for personnel working at the plant. This summary focuses on describing the work performed for the in-vessel shielding design and analysis aiming at limiting the secondary sodium activation.

Fei, T.↗

DOE OSTI2022

Boron injection system for controlling a nuclear reaction by delivering boron into a containment vessel

A nuclear power system includes a reactor vessel that includes a reactor core mounted within a volume of the reactor vessel. The reactor core includes one or more nuclear fuel assemblies configured to generate a nuclear fission reaction. The nuclear power system further includes a containment vessel sized to enclose the reactor vessel such that an open volume is defined between the containment vessel and the reactor vessel. A boron injection system is positioned in the open volume of the containment vessel and includes an amount of boron sufficient to stop the nuclear fission reaction or maintain the nuclear fission reaction at a sub-critical state. The boron injection system is positioned to deliver the amount of boron into the open volume.

Linik, Larry↗

DOE OSTI2022

Heat transfer systems for nuclear reactor cores, and related systems

A system for transferring heat from a nuclear reactor comprises a nuclear reactor comprising a nuclear fuel and a reactor vessel surrounding the nuclear reactor and a heat transfer system surrounding the nuclear reactor. The heat transfer system comprises an inner wall surrounding the nuclear reactor vessel, first fins coupled to an outer surface of inner wall, an outer wall between the inner wall and a surrounding environment, and second fins coupled to an inner surface of the outer wall and extending in a volume between the outer surface of the inner wall and the inner surface of the outer wall, the outer surface of the inner wall and the first fins configured to transfer heat from the nuclear reactor core to the second fins and the inner surface of the outer wall by thermal radiation. The heat transfer system may be directly coupled to the nuclear reactor vessel, or may be coupled to an external reflector surrounding the nuclear reactor vessel. Related heat transfer systems and systems for selectively removing heat from a nuclear reactor are disclosed.

Ougouag, Abderrafi Mohammed El-Amine↗

DOE OSTI2016

SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. To use Grizzly for probabilistic analysis, it is necessary to have a way to rapidly evaluate stress intensity factors. To accomplish this goal, a reduced order model (ROM) has been developed to efficiently represent the behavior of a detailed 3D Grizzly model used to calculate fracture parameters. This approach uses the stress intensity factor influence coefficient method that has been used with great success in FAVOR. Instead of interpolating between tabulated solutions, as FAVOR does, the ROM approach uses a response surface methodology to compute fracture solutions based on a sampled set of results used to train the ROM. The main advantages of this approach are that the process of generating the training data can be fully automated, and the procedure can be readily used to consider more general flaw configurations. This paper demonstrates the procedure used to generate a ROM to rapidly compute stress intensity factors for axis-aligned flaws. The results from this procedure are in good agreement with those produced using the traditional influence coefficient interpolation procedure, which gives confidence in this method. This paves the way for applying this procedure for more general flaw configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Reactor Pressure Vessel Characterization Methods Development for the High Flux Isotope Reactor to Support Future Vessel Replacement

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Chandler, David↗

DOE OSTI2015

Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

PWR decay heat removal system in which steam from the pressurizer drives a turbine which drives a pump to inject water into the reactor pressure vessel

In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

PWR decay heat removal system in which steam from the pressurizer drives a turbine which drives a pump to inject water into the reactor pressure vessel

Graham, Thomas G.↗

DOE OSTI2014

Reactor pressure vessel head vents and methods of using the same

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Graphite Research Reactor (RFT) Spectrometric Investigation During Dismantling - 17006

Dismantling of research reactors are carried out at the National Research Centre 'Kurchatov Institute'. Technical start-up of the 20 MW capacity experimental graphite water cooling reactor (RFT) was started up in 1952. During experimental work, some accidents, followed by a pollution of a graphite blocks by fission products of nuclear fuel, took place. After 10 years working period the reactor was shut down. The fuel was removed and the reactor was partially dismantled. Reactor graphite stack (cylinder 380 cm high and 260 cm in diameter) was isolated. RFT reactor dismantling was began in 2015. Video investigation of reactor channels in graphite stack was made before dismantling. Graphite in channels near active zone of reactor was destroyed. Measurements of exposure dose rate inside the channels near active zone showed value up to 2 Sv. The samples were taken from different layers of stack during graphite blocks removing from the reactor. Complex α-, β-, γ- spectrometric analysis of this samples was made. The graphite blocks were contaminated by fuel components and its fission products. It was found that {sup 137}Cs and {sup 90}Sr were the main contamination radionuclides in graphite. After graphite block removing and reactor vessel bottom cleaning, exposure dose rate reached 5-7 mSv/h at upper level of the reactor vessel and at the bottom ∼40 mSv/h at the bottom. Gamma-spectrometric analysis of reactor vessel was made. These measurements showed that activated in iron reactor vessel {sup 60}Co made the main contribution to exposure dose rate. After reactor vessel dismantling exposure dose rate reached 2 mSv/h at upper level of the reactor vessel and∼ 3 mSv/h at the bottom. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels

Embrittlement of light water reactor pressure vessels (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. In this study, the steels were irradiated to a fluence of ~ 1.38 x 10 20 n/cm 2 at ~ 292 °C in the UCSB ATR-2 experiment. In Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promotes precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Vibratory Processes in Vessel Steel Damage Kinetics in Nuclear Reactors

The possible reasons for nonstandard embrittlement and significant variance of the experimental data in the study of the impact of neutron irradiation on the properties of reactor vessel materials are analyzed. The authors assume on the basis of experimental data and analysis of foreign research that the fluence dependence of the degree of radiation embrittlement can be of nonmonotonic as well as cyclic. The observation of vibratory processes in the embrittlement kinetics of vessel steel attests periodic transformation of structure and change in the properties of the metal. It is expected that the defect structure formed as a result of neutron bombardment undergoes transformation once or periodically in the direction of losses as well as restoration of the initial properties. The mentioned data shape the search for methods of regulating the radiation resistance of materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a HTGR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR ha s undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. The generic plant model is based on publicly available PMBR-400 design information. For plant aspects (e.g., reactor building size and leak rate) that are not described in the PBMR-400 references, the analysts made assumptions needed to construct a MELCOR full-plant model. The HTGR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to a recuperative heat exchange r. Surrounding the reactor vessel is a reactor cavity contained within a confinement room cooled by the Reactor Cavity Cooling System (RCCS). Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2018

Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the Grizzly code

In light water reactor (LWR) nuclear power plants, the reactor pressure vessel (RPV) plays an essential safety role, and its integrity must be ensured during a variety of transient loading conditions. These can include off-normal conditions such as a pressurized thermal shock (PTS), as well as transients encountered during normal startup, shutdown, and testing of the reactor. Exposure to irradiation and elevated temperatures embrittles the RPV’s steel over time, making it increasingly susceptible to failure due to propagation of fractures that could initiate at the locations of flaws introduced during the manufacturing process. As long-term operation scenarios are being considered for LWRs in the United States, it is important to have a flexible simulation tool that can be used to perform probabilistic evaluations of RPV integrity under a wide variety of conditions and incorporate improved predictive models of RPV steel embrittlement. The Grizzly code is being developed to meet these needs. This paper describes Grizzly’s modular architecture, provides results of benchmarking studies of various components of Grizzly, and demonstrates the application of Grizzly on a model that includes plume effects that are difficult to represent in other codes being used in current practice.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗