Graphite-moderated Nuclear Reactors Research Articles

Generic method to assess transmutation feasibility for nuclear waste treatment and application to irradiated graphite

Graphite-moderated nuclear reactors have already produced more than 250,000 tons of irradiated nuclear graphite, or i-graphite, world-wide. The sustainability of this technology relies on the end-of-life management of its moderator, which is activated into a long-lived, low or intermediate-level nuclear waste, by neutron fluxes, during operating time. In particular, carbon-14 is created. Nuclear transmutation, enabled by laser-driven particle acceleration, has been envisioned as a potential novel treatment scheme for long-lived nuclear waste. By triggering controlled nuclear reactions with energetic particles, long-lived radio-nuclides could be transformed into short-lived or stable isotopes. Such a system could treat the carbon-14 nuclei trapped within the i-graphite matrix, an isotope which is difficult to isolate by other means. This work performs a quantitative preliminary study of this transmutation scheme, in order to assess its feasibility at an industrial scale. The method used can be transposed to assess any transmutation scheme using a beam of particles directly sent on the material to be treated. First, a nuclear interaction channel which transmutes carbon-14 nuclei without creating new long-lived radio-nuclides is identified. It consists in the choice of a type of particle, among which protons, γ photons and neutrons can all be accelerated by laser–matter interaction; and it is completed by the adequate energy at which this particle must be sent on i-graphite. To that end, the nuclear cross-sections of carbon-12, carbon-13 and carbon-14 are reviewed, neglecting other impurities in i-graphite. Then, based on the interaction channel identification, the energy cost of this scheme is estimated. Protons between 1 and 5MeV make it possible to transmute carbon-14 without creating any new long-lived activity. However, our result show that, even in this favourable reaction channel, the transmutation energy cost is too high for an i-graphite transmutation scheme to be industrially feasible.

Determination of nuclear graphite impurities by prompt gamma activation analysis to support decommissioning operations

Elemental composition of non-irradiated nuclear grade graphite was determined by Prompt Gamma Activation Analysis (PGAA) to quantitively assess some neutron activation precursors. The knowledge of these data is paramount for an accurate radiological characterization of the material before decommissioning of graphite-moderated nuclear reactors. Bulk results of most of the nuclides were consistent with benchmark Inductively Coupled Plasma–Mass Spectrometry (ICP-MS) and literature data. In the case of 14N and 35Cl, depth distribution profiles were observed and quantified for the first time. These outcomes could shine a light on 14C and 36Cl depth distribution profiles and fractional release during leaching experiments.

Computational fluid dynamics assessment of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) system for the Transient REActor Test (TREAT) facility

The Transient REActor Test facility (TREAT) is an air-cooled, graphite moderated nuclear reactor that is used for irradiation testing of nuclear fuels and materials. The shutdown of the Halden reactor in Norway lead to an increased interest of an indigenous nuclear facility that can perform Reactivity Initiated Accident (RIA) tests for Light Water Reactors (LWRs). Idaho National Laboratory (INL) has been leading the effort of making minor modifications to the TREAT facility to enhance its testing capabilities so it can mimic the neutronic boundary conditions representative of a RIA for a LWR. The power burst of the TREAT reactor needs to be shortened from 89 to 40 msec. to properly irradiate the fuel sample under RIA conditions. INL has proposed to use helium-3, a strong neutron absorber, as a form of negative reactivity to shorten the pulse of the reactor. In support of the development of a transient control rod for the TREAT, a Computational Fluid Dynamics (CFD) model is needed to confidently simulate the density evolution of helium-3 in this system.

Mimicking irradiation-induced cracking of nuclear graphite using bromine intercalation

Keyway root cracking and radiolytic weight loss in graphite moderated nuclear reactors may limit future successful operation of the civil nuclear fleet. A novel means of initiating fracture using internally generated stresses has been developed, which uses bromine intercalation to induce fracture in small graphite samples. Successful crack initiation and propagation has been achieved in 1:10 scale reactor components. In situ X-ray Computed Tomography during bromination has enabled observations of failure in real time, allowing the role of microstructure during fracture and weight loss to be determined. Crack paths predicted by models have been assessed.

Mimicking Irradiation-Induced Cracking of Nuclear Graphite Using Bromine Intercalation

Keyway root cracking and radiolytic weight loss in graphite moderated nuclear reactors may limit future successful operation of the civil nuclear fleet. A novel means of initiating fracture using internally generated stresses has been developed, which uses bromine intercalation to induce fracture in small graphite samples. Successful crack initiation and propagation has been achieved in 1:10 scale reactor components. In situ X-ray Computed Tomography during bromination has enabled observations of failure in real time, allowing the role of microstructure during fracture and weight loss to be determined. Crack paths predicted by models have been assessed.

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥800 °C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

Assessing the effectiveness of a plant modification from a small set of degradation endpoints: Predicting the significance of evidence from an ongoing experiment

The present study aims to establish how much information, in the form of the endpoints of a physical process that is currently ongoing, is needed to evaluate the success of a tentative plant modification aimed at reducing some type of degradation. This will make possible a timely decision on the enhancement, continuation or rollback of the modification.The irradiation dependency of the amount of carbonaceous deposit observed on fuel discharged from six graphite-moderated nuclear reactors is analyzed by a two-stage Chan-Karolyi-Longstaff-Sanders process. A novel heuristic technique is introduced for the fitting of the parameters of a stochastic differential model to a set of path endpoints. The model is altered to simulate a reduction in the rate of carbon deposition on newly loaded fuel, caused by a targeted change in the reactor coolant chemistry. Synthetic sets of carbon deposit thickness readings in highly irradiated fuel are generated. The statistical significance of the discrepancy between the distributions of the experimental and synthetic data is then quantified as a function of the simulated effectiveness of the coolant chemistry change and the size of the synthetic data set.

LIMITED OXIDATION OF IRRADIATED GRAPHITE WASTE TO REMOVE SURFACE CARBON-14

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ( 14 C), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the 14 C, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create COx gases, i.e. “gasify” graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl- like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide (CO₂). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

Temperature monitoring of nuclear reactor cores with multiplexed fiber Bragg grating sensors

In-core temperature measurement is a critical issue for the safe operation of nuclear reactors. Classical thermocouples require shielded connections and are known to drift under high neutron fluence. As an alternative, we propose to take advantage of the multiplexing ca- pabilities of fiber Bragg grating (FBG) temperature sensors. Our experi- ments show that sensitivity to radiation depends on both the radiation field and the grating characteristics. For some FBGs installed in an air- cooled graphite-moderated nuclear reactor the difference between the measurements and the readings of calibrated backup thermocouples was within the measurement uncertainty. In the worst case, the differ- ence saturated after 30 h of reactor operation at about 5°C. To reach megagray per hour level gamma-dose rates and 10 19 neutron/cm 2 flu- ences, we irradiated multiplexed FBG sensors in a material testing nuclear reactor. At room temperature, FBG temperature sensors can sur- vive in such radiation conditions, but at 90°C a severe degradation is observed. We evidence the possibility to use FBG sensing technology for in-core monitoring of nuclear reactors with specific care under well- specified conditions. © 2002 Society of Photo-Optical Instrumentation Engineers.

Pact seeks to end North Korean nuclear threat

A U.S.-North Korean agreement signed Aug. 12 takes a major step toward allaying world concern about North Korea's suspected effort to develop nuclear arms. Under the pact, North Korea has agreed to replace its graphite-moderated nuclear reactors with Western-style light-water reactors. In turn, the U.S. will try to arrange financing for the light-water reactors and help provide interim power sources during their construction. North Korea's current three graphite-moderated reactors—a 5-MW one already in operation and 50-MW and 200-MW ones under contruction—are ideally suited for making weapons-grade plutonium. Light-water reactors also make plutonium, but produce less per unit of energy generated. And their plutonium, although usable in weapons, makes less efficient arms than does plutonium from graphite-moderated reactors. The accord calls for North Korea to freeze construction on the 50-MW and 200-MW reactors, seal its radiochemical lab that studies nuclear reprocessing, and shut down its 5-MW reactor—af...

Design and Economic Study of a Gas-Phase Chemonuclear Ozone Process

An evaluation of the economics of producing ozone from oxygen in a system utilizing fission fragment energy is presented with a study of the design of such a system. The study covers a range of G values for ozone formation from 3 to 15 and chemonuclear reactor ozone concentrations from 10 parts/106 to 10% at an operating temperature of −20°C. A graphite-moderated nuclear reactor design, which utilizes a 2.5-µm thick U-Pd alloy foil as fuel elements, is employed. Ozone is separated by silica-gel adsorption; decontamination procedures are described. Investment and operating costs for the chemonuclear and conventional methods of ozone production are compared for an ozone production rate of 100 ton/d. It is concluded that the chemonuclear route becomes competitive with conventional ozonizers at a steady-state concentration of 0.1% ozone for a G value of 9 and at a concentration as low as 150 parts/106 for a G value of 15. The study indicates the need for research on fission fragment and radiation chemistry in the ozone-oxygen system at temperatures from 20°C to −78°C for determining the feasibility of these yields.

Hallam Nuclear Power Facility Electric Auxiliary-Power Systems

This paper describes the electric auxiliary-power systems in a sodium-cooled graphite-moderated nuclear reactor (SGR). Reliability requirements and methods of achieving them are discussed, and sodium-system preheating, a requirement unique to sodium-cooled reactors, is outlined. Nuclear instrumentation electric noise problems and methods of avoiding them are included. Auxiliary power requirements of a 360-mwe (electrical megawatt) SGR plant are compared with those of a similar conventional plant.

Hallam Nuclear Power Facility Prototype for Advanced Sodium-Cooled Power Stations

The sodium-cooled graphitemoderated nuclear reactor concept is described. The Hallam Nuclear Power Facility (HNPF) for producing superheated steam to generate 75 electrical megawatts (mwe) net power represents a significant step toward the ultimate capabilities of this concept. A larger plant design, capable of 360-mwe gross output, is also presented. The advanced design is expected to produce steam thermodynamically comparable to modern steam plants economically competitive with fossil fueled stations in many areas of the United States. Two important characteristics of the sodium-cooled reactor concept are the excellent heat ransfer properties of liquid sodium and the inherent safety of a low-pressure reactor system. These characteristics are utilized extensively in this concept to minimize total power cost.

Gas Content of Graphites

ARTIFICIAL graphite for use in graphite-moderated nuclear reactors, as anodes for mercury vapour rectifiers, and crucibles for vacuum fusion work should contain little gas. Otherwise, when the graphite is heated to operating temperatures, the gas is desorbed, resulting in adverse effects. For example, in the graphite-moderated nuclear reactor, hydrogen and carbon monoxide released from the graphite can interact with the fuel-cladding material (usually zirconium or stainless steel), resulting in hardening and embrittlement of the metal. Released carbon monoxide can disproportionate on cooler metal heat-exchanger surfaces depositing carbon, resulting in an adverse effect on the heat-transfer and flow properties of the system.