Accelerator Production Of Tritium Research Articles

Accelerators for Fusion Materials Testing

Fusion materials research is a worldwide endeavor as old as the parallel one working toward the long term stable confinement of ignited plasma. In a fusion reactor, the preservation of the required minimum thermomechanical properties of the in-vessel components exposed to the severe irradiation and heat flux conditions is an indispensable factor for safe operation; it is also an essential goal for the economic viability of fusion. Energy from fusion power will be extracted from the 14 MeV neutron freed as a product of the deuterium–tritium fusion reactions; thus, this kinetic energy must be absorbed and efficiently evacuated and electricity eventually generated by the conventional methods of a thermal power plant. Worldwide technological efforts to understand the degradation of materials exposed to 14 MeV neutron fluxes [Formula: see text] m[Formula: see text]s[Formula: see text], as expected in future fusion power plants, have been intense over the last four decades. Existing neutron sources can reach suitable dpa (“displacement-per-atom”, the figure of merit to assess materials degradation from being exposed to neutron irradiation), but the differences in the neutron spectrum of fission reactors and spallation sources do not allow one to unravel the physics and to anticipate the degradation of materials exposed to fusion neutrons. Fusion irradiation conditions can be achieved through Li (d, xn) nuclear reactions with suitable deuteron beam current and energy, and an adequate flowing lithium screen. This idea triggered in the late 1970s at Los Alamos National Laboratory (LANL) a campaign working toward the feasibility of continuous wave (CW) high current linacs framed by the Fusion Materials Irradiation Test (FMIT) project. These efforts continued with the Low Energy Demonstrating Accelerator (LEDA) (a validating prototype of the canceled Accelerator Production of Tritium (APT) project), which was proposed in 2002 to the fusion community as a 6.7[Formula: see text]MeV, 100[Formula: see text]mA CW beam injector for a Li (d, xn) source to bridge with the International Fusion Materials Irradiation Facility (IFMIF) under discussion at the time. Worldwide technological efforts are maturing soundly and the time for a fusion-relevant neutron source has arrived according to world fusion roadmaps; if decisions are taken we could count the next decade with a powerful source of 14 MeV neutrons thanks to the expected significant results of the Engineering Validation and Engineering Design Activity (EVEDA) phase of the IFMIF project. The accelerator know-how has matured in all possible aspects since the times of FMIT conception in the 1970s; today, operating 125 mA deuteron beam at 40 MeV in CW with high availabilities seems feasible thanks to the understanding of the beam halo physics and the three main technological breakthroughs in accelerator technology: (1) the ECR ion source for light ions developed at Chalk River Laboratories in the early 1990s, (2) the RFQ operation of H[Formula: see text] in CW with 100 mA demonstrated by LEDA in LANL in the late 1990s, and (3) the growing maturity of superconducting resonators for light hadrons and low [Formula: see text] beams achieved in recent years.

Tritium supply assessment for ITER and DEMOnstration power plant

The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

Performance of a Clad Tungsten Rod Spallation Neutron Source Target

Tungsten rods, slip-clad with Type 304L stainless steel, performed successfully as a spallation neutron source target operating to a peak fluence of ~4 × 1021 p/cm2. The target was used as a neutron source during the Accelerator Production of Tritium (APT) materials irradiation program at the Los Alamos Neutron Science Center. Tungsten rods of 2.642-mm diameter were slip-fit in Type 304L stainless steel tubes that had an inner diameter of 2.667 mm. The radial gap was filled with helium at atmospheric pressure and room temperature. Los Alamos High Energy Transport (LAHET) calculations suggest a time-averaged peak power deposition in the W of 2.25 kW/cm3. Thermal-hydraulic calculations indicate that the peak centerline W temperature reached 271°C. The LAHET calculations were also used to predict neutron and proton fluxes and spectra for the complex geometry used in the irradiation program. Activation foil sets distributed throughout the experiment were used to determine target neutronics performance as a comparison to the LAHET calculations. Examination of the irradiated target assemblies revealed no significant surface degradation or corrosion on either the Type 304L or the W surfaces. However, it was clear that the irradiation changed material properties because post-proton-irradiation measurements on Type 304L test samples from the APT program demonstrated increases in the yield strength and decreases in the ductility and fracture toughness with increasing dose, and the wrought W rod samples became brittle. Fortunately, the slip-clad target design subjects the materials to very low stress.

The effect of 800 MeV proton irradiation on the mechanical properties of tungsten at room temperature and at 475 °C

For the accelerator production of tritium (APT), the accelerator driven transmutation facility (ADTF), and the advanced fuel cycle initiative (AFCI), tungsten is being proposed as a target material to produce neutrons. In this study, tungsten rods were irradiated at the 800 MeV Los Alamos Neutron Science Center (LANSCE) proton accelerator for six months. After irradiation to a maximum dose in the tungsten of 23.3 dpa at T irr = 50–270 °C, the rods were sliced into sections, hardness tests were performed at room temperature, and compression tests were performed at room temperature and at 475 °C to assess the effect of irradiation on the mechanical properties of tungsten. The results show an increase in strength and a decrease in ductility with dose. Specimens tested at 475 °C had lower yield strength and reduced work hardening capability compared to specimens tested at room temperature.

The effect of coatings on retention and permeation in SS 316L APT tritium production tubes

The accelerator production of tritium (APT) design calls for thousands of thin-walled tubes to be filled with 3He gas. The reaction of the spallation neutrons with this gas will result in the bombardment of the interior of these tubes with energetic tritons and protons. For APT to be able to meet its tritium production goals, it is necessary that the holdup of the tritium in the tube walls be minimized. To examine the tritium retention characteristics of stainless steel, one of the tube reference materials, accelerator implantation experiments were performed. In these experiments, deuterium was used in place of tritium to eliminate the problem of tritium contamination. Deuterons with energies up to 200 keV and protons with energies up to 600 keV were implanted into stainless steel (SS 316L) samples to fluences as high as 5 × 10 22 D/m 2 and 5 × 10 22 p/m 2. Thermal desorption spectroscopy showed that approximately 3% of the deuterium was retained within the sample. Approximately 0.5% of the deuterium permeated through to the back surface of the sample where a zirconium getter trapped the deuterium. The deuterium trapped in the zirconium layer was measured by nuclear reaction analysis. Eight-micron thick copper and nickel coatings were applied to the implantation side of the stainless steel substrate in an attempt to reduce the retention and permeation of the deuterium. The copper-coated stainless steel was not successful in reducing the retention and permeation, however the nickel coated stainless steel reduced both the retention and permeation substantially.

Hydrogen release from 800 MeV proton-irradiated tungsten

Tungsten irradiated in spallation neutron sources, such as those proposed for the accelerator production of tritium (APT) project, will contain large quantities of generated helium and hydrogen gas. Tungsten used in proposed fusion reactors will also be exposed to neutrons, and the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. The release kinetics of these gases during various off-normal scenarios involving loss of coolant and after heat-induced rises in temperature are of particular interest for both applications. To determine the release kinetics of hydrogen from tungsten, tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANSCE) to high exposures as part of the APT project have been examined. Hydrogen evolution from the tungsten has been measured using a dedicated mass-spectrometer system by subjecting the specimens to an essentially linear temperature ramp from ∼300 to ∼1500 K. Release profiles are compared with predictions obtained using the Tritium Migration Analysis Program (TMAP4). The measurements show that for high proton doses, the majority of the hydrogen is released gradually, starting at about 900 K and reaching a maximum at about 1400 K, where it drops fairly rapidly. Comparisons with TMAP show quite reasonable agreement using a trap energy of 1.4 eV and a trap density of ∼7%. There is a small additional release fraction occurring at ∼550 K, which is believed to be associated with low-energy trapping at or near the surface, and, therefore, was not included in the bulk TMAP model.

The Effect of 800 MeV Proton Irradiation on the Mechanical Properties of Tungsten

For the Accelerator Production of Tritium (APT) and the Accelerator Driven Transmutation Facility (ADTF), tungsten is being proposed as a target material to produce neutrons. Previous work has shown that the mechanical properties of tungsten are degraded from irradiation in a fission neutron flux but little work has been performed on the irradiation of tungsten in a high energy proton beam. In this study, tungsten rods were irradiated at the 800 MeV Los Alamos Neutron Science Center (LANSCE) proton accelerator for six months. To avoid corrosion during irradiation, the rods were slip fit with thin (0.25 mm thick) 304L stainless steel (SS) or (0.125 mm thick) annealed Alloy 718 tubing. After irradiation to a maximum dose in the tungsten of 23.3 dpa at Tirr = 50‐270 ◦ C, the clad rods were opened in the hot cells and the tungsten was removed. The tungsten was then sliced into short compression specimens (∼ 3 mm long). Hardness tests and compression tests were performed on the tungsten rods to assess the effect of irradiation on their mechanical properties. Results show an increase in hardness with dose and irradiation temperature and an increase in yield stress with dose.

The mechanical properties of 316L/304L stainless steels, Alloy 718 and Mod 9Cr–1Mo after irradiation in a spallation environment

The Accelerator Production of Tritium (APT) project proposes to use a 1.0 GeV, 100 mA proton beam to produce neutrons via spallation reactions in a tungsten target. The neutrons are multiplied and moderated in a lead/aluminum/water blanket and then captured in 3 He to form tritium. The materials in the target and blanket region are exposed to protons and neutrons with energies into the GeV range. The effect of irradiation on the tensile and fracture toughness properties of candidate APT materials, 316L and 304L stainless steel (annealed), modified (Mod) 9Cr–1Mo steel, and Alloy 718 (precipitation hardened), was measured on tensile and fracture toughness specimens irradiated at the Los Alamos Neutron Science Center accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The irradiation temperatures ranged from 50°C to 164°C, prototypic of those expected in the APT target/blanket. The maximum achieved proton fluence was 4.5×10 21 p/ cm 2 for the materials in the center of the beam. This maximum exposure translates to a dpa of 12 and the generation of 10 000 appm H and 1000 appm He for the Type 304L stainless steel tensile specimens. Specimens were tested at the irradiation temperature of 50–164°C. Less than 1 dpa of exposure reduced the uniform elongation of the Alloy 718 (precipitation hardened) and Mod 9Cr–1Mo to less than 2%. This same dose reduced the fracture toughness by 50%. Approximately 4 dpa of exposure was required to reduce the uniform elongation of the austenitic stainless steels (304L and 316L) to less than 2%. The yield stress of the austenitic steels increased to more than twice its non-irradiated value after less than 1 dpa. The fracture toughness reduced significantly by 4 dpa to ∼100 MPa m 1/2. These results are discussed and compared with results of similar materials irradiated in fission reactor environments.

Determination of helium and hydrogen yield from measurements on pure metals and alloys irradiated by mixed high energy proton and spallation neutron spectra in LANSCE

The confident design of accelerator-driven spallation neutron devices will require good estimates of the cross-sections for generation of helium and hydrogen in the mixed spectra of high energy protons and neutrons that will be experienced by the structural materials. Improved estimates of these cross-sections were derived from a series of irradiations that were conducted at relatively low temperatures (<100°C) in the Los Alamos Neutron Science Center (LANSCE) as part of the test program supporting the Accelerator Production of Tritium (APT) Program. In this irradiation campaign, a variety of candidate structural alloys and pure metal dosimeter foils were irradiated in various particle spectra, ranging from 800 MeV protons, to mixed energy distributions of both protons and spallation neutrons, and finally to distributions consisting primarily of high energy neutrons. At proton energies on the order of hundreds of MeV, exceptionally high levels of gas atoms are generated in all elemental constituents of typical iron-based and nickel-based structural alloys, with helium typically on the order of ∼150 appm per dpa and hydrogen at approximately a factor of 3–5 higher. Most of the gas production is due to proton and helium recoils from the proton beam interactions with the specimens, although gas and especially damage production from lower-energy spallation neutrons becomes increasingly significant at locations farther from the beam center. The results show that helium production rate per dpa by protons in elements typically found in structural alloys is relatively insensitive to elemental composition. The measured helium concentrations and the derived cross-sections are larger by about a factor of two, however, than those calculated using the LAHET code which was optimized for prediction of neutron/proton ratios in the target tungsten source rods of the APT test. Unlike helium, the retained hydrogen levels are somewhat sensitive to composition, reflecting primarily different levels of diffusional loss, but hydrogen is still retained at rather high concentrations, allowing a lower bound estimate of the hydrogen generation rates.

Microstructural origins of radiation-induced changes in mechanical properties of 316 L and 304 L austenitic stainless steels irradiated with mixed spectra of high-energy protons and spallation neutrons

A number of candidate alloys were exposed to a particle flux and spectrum at Los Alamos Neutron Science Center (LANSCE) that closely match the mixed high-energy proton/neutron spectra expected in accelerator production of tritium (APT) window and blanket applications. Austenitic stainless steels 316 L and 304 L are two of these candidate alloys possessing attractive strength and corrosion resistance for APT applications. This paper describes the dose dependence of the irradiation-induced microstructural evolution of SS 316 L and 304 L in the temperature range 30–60°C and consequent changes in mechanical properties. It was observed that the microstructural evolution during irradiation was essentially identical in the two alloys, a behavior mirrored in their changes in mechanical properties. With one expection, it was possible to correlate all changes in mechanical properties with visible microstructural features. A late-term second abrupt decrease in uniform elongation was not associated with visible microstructure, but is postulated to be a consequence of large levels of retained hydrogen measured in the specimens. In spite of large amounts of both helium and hydrogen retained, approaching 1 at.% at the highest exposures, no visible cavities were formed, indicating that the gas atoms were either in solution or in subresolvable clusters.

Correlation of radiation-induced changes in mechanical properties and microstructural development of Alloy 718 irradiated with mixed spectra of high-energy protons and spallation neutrons

Alloy 718 is a γ ′(Ni 3(Al,Ti))–γ″(Ni 3Nb) hardenable superalloy with attractive strength, and corrosion resistance. This alloy is a candidate material for use in accelerator production of tritium (APT) target and blanket applications, where it would have to withstand low-temperature irradiation by high-energy protons and spallation neutrons. The existing data base, relevant to such irradiation conditions, is very limited. Alloy 718 has therefore been exposed to a particle flux and spectrum at the Los Alamos Neutron Science Center (LANSCE), closely matching those expected in the APT target and blanket applications. The yield stress of Alloy 718 increases with increasing dose up to ∼0.5 dpa , and then decreases with further increase in dose. The uniform elongation, however, drastically decreases with increasing dose at very low doses (<0.5 dpa), and does not recover when the alloy later softens somewhat. Transmission electron microscopy (TEM) investigation of Alloy 718 shows that superlattice spots corresponding to the age-hardening precipitate phases γ ′ and γ″ are lost from the diffraction patterns for Alloy 718 by only 0.6 dpa, the lowest proton-induced dose level achieved in this experiment. Examination of samples that were neutron irradiated to doses of only ∼0.1 dpa showed that precipitates are faintly visible in diffraction patterns but are rapidly becoming invisible. It is proposed that the γ ′ and γ″ first become disordered (by <0.6 dpa), but remain as solute-rich aggregates that still contribute to the hardness at relatively low dpa levels, and then are gradually dispersed at higher doses.

Thermal-Hydraulic Design of the Accelerator Production of Tritium Tungsten Neutron Source

The thermal-hydraulic design of the accelerator production of tritium (APT) tungsten neutron source is presented. A carefully engineered thermal-hydraulic design is required to remove the deposited power effectively during normal operations and remove the decay power during plant shutdown and postulated accidents. For steady-state operations and operational and anticipated transients, the design criterion is to maintain single-phase flow conditions with a margin to onset of nucleate boiling. The margin is determined based on phenomenological and geometric uncertainties associated with the design. A large margin to thermal excursion limits, such as critical heat flux and onset of flow instability, also is maintained during normal operations. In general, a very robust thermal-hydraulic design can be accomplished using the traditional models and correlations available in the engineering literature. However, two issues require further attention: maintaining adequate flows in a parallel network of flow channels and minimizing the volume fraction of heavy water to maximize tritium production.The design uses ladderlike structures that contain clad tungsten cylinders in the rungs that have coolant supplied and removed by the vertical ladder rails. Because the power density drops in the beam direction, the thickness of the tungsten cylinders is increased with increasing beam penetration length. The cooling requirement is determined using a conservative criterion where the minimum wall subcooling inside the rungs is at least 40°C and the minimum Reynolds number is 6000. Initial flow distribution tests were conducted with a full-scale model of an APT ladder assembly based on a preliminary design. Flow distributions can be made more even by using a larger riser than downcomer and also by increasing the flow resistance across each rung. The calculations discussed assume nominal dimensions, even though the power deposition and removal use a conservative approach. The effect of manufacturing tolerances will be investigated in future research. Also, the applicability of the critical heat flux and onset of flow instability models to small coolant channels is being verified experimentally. Further design optimization will be possible when these studies are completed.

The Accelerator Production of Tritium Materials Test Program

A materials qualification program has been developed to irradiate and test candidate materials (alloy 718, Type 316L, and Type 304L stainless steel, modified Fe9Cr-1Mo(T91), Al-6061-T6, and Al-5052-O) for use in the Accelerator Production of Tritium (APT) target and blanket. The irradiations were performed in prototypic proton and neutron spectra at prototypic temperatures (50 to 160°C). The study used the 800-MeV, 1.0-mA proton accelerator at the Los Alamos Neutron Science Center, which produces a Gaussian beam with 2 sigma = 3 cm. The experiment geometry is arranged to contain near-prototypic modules of the tungsten neutron source and the lead and aluminum blanket as well as mechanical test specimens of candidate APT materials. The particle spectrum varies throughout the irradiation volume; specimens are exposed to protons and a variety of mixed proton and neutron spectra, depending on the specimen’s position relative to the beam center. These specimens have been irradiated for >3600 h to a maximum proton fluence of 4 × 1021 p/cm2 in the center of the proton beam. Specimens will yield data on the effect of proton irradiation, to high dose, on material properties from tensile tests, three-point bend tests, fracture toughness tests, pressurized tubes, U-bend stress corrosion cracking specimens, corrosion measurements, and microstructural characterization using transmission electron microscopy specimens. Results from these studies are applicable to all spallation neutron sources now in operation and under consideration, including the Spallation Neutron Source, the European Spallation Source, and The Accelerator Transmutation of Waste project.

An Object-Oriented Parallel Particle-in-Cell Code for Beam Dynamics Simulation in Linear Accelerators

We present an object-oriented three-dimensional parallel particle-in-cell (PIC) code for simulation of beam dynamics in linear accelerators (linacs). An important feature of this code is the use of split-operator methods to integrate single-particle magnetic optics techniques with parallel PIC techniques. By choosing a splitting scheme that separates the self-fields from the complicated externally applied fields, we are able to utilize a large step size and still retain high accuracy. The method employed is symplectic and can be generalized to arbitrarily high order accuracy if desired. A two-dimensional parallel domain decomposition approach is employed within a message-passing programming paradigm along with a dynamic load balancing scheme. Performance tests on an SGI/Cray T3E-900 and an SGI Origin 2000 show good scalability of the object-oriented code. We present, as an example, a simulation of high current beam transport in the accelerator production of tritium (APT) linac design.

Residual Stresses in an Inconel-718-Clad Tungsten Tube Processed by Hot Iso-static Pressing

We have completed a comparative X-ray diffraction (XRD) and Finite Element Analysis (FEA) study of Inconel 718 clad Tungsten tubes, of relevance to the Accelerator Production of Tritium (APT) program at the Los Alamos National Laboratory (LANL). Because the Inconel cladding must be in intimate contact with the tungsten at operating temperatures, the materials are diffusion bonded by Hot Iso-static Pressing (HIP) at 1080°C. However, the difference in thermal expansion coefficients of inconel 718 and tungsten produce large residual stresses upon cooling. The classical XRD (d vs. sin 2 ψ) technique was used to determine the stresses in the outer layer of inconel 718. These results are used to benchmark Finite Element Analysis (FEA) calculations, and a direct comparison is made in this paper.

Helium/Hydrogen Effects on the Properties of Materials for the APT Target/Blanket Region

Materials irradiated in the target/blanket region of the Accelerator Production of Tritium (APT) system will contain significantly more hydrogen and helium, for any given displacements per atom (dpa), than typically found in materials irradiated in a fission reactor. The individual effects of helium and hydrogen implantation and displacement damage on structural materials are relatively well established. Helium will increase the strength, decrease the ductility, reduce the creep and stress rupture properties, decrease the fatigue life and weldability, and promote swelling. Hydrogen will also adversely effect the mechanical properties and promote premature fracture along metallurgical interfaces. Displacement damage generally increases the strength and decreases the ductility of single-phase metals and alloys. The adverse effects of hydrogen and helium implantation may couple with displacement damage to limit component lifetime in the target/blanket region of the APT system. This paper provides a technical basis to rationalize potential synergistic effects among displacement damage and the hydrogen and helium embrittlement processes and suggests that such synergistic effects may be of significant importance to component performance in intense spallation neutron sources.

Design and Operation of the APT Target/Blanket

The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of the T/B assembly and the attendant heat-removal systems. The T/B assembly produces tritium using a high-energy proton beam and a spallation neutron source. All systems reside within the T/B building, located at the end of a linear accelerator. During nominal operation, protons are accelerated to an energy of 1030MeV at a current of 100mA. The protons interact with tungsten and lead nuclei in the T/B assembly to produce neutrons through the process of nuclear spallation. Neutron capture in 3He gas produces tritium, which is removed on a semicontinuous basis. The T/B assembly is modular, to allow for replacement of spent components and minimization of waste. Inconel and Aluminum structural materials provide containment of spallation products, a coolant pressure boundary, and corrosion resistance. The operating environment is a combination of high-energy protons and neutrons. All structures operate at low temperature and are cooled with water. Structures directly in the proton beam, which endure the highest irradiation damage, are replaced annually. Materials in the blanket region are primarily in a low-neutron flux environment and potentially will last the life of the facility. Published by Elsevier Science Inc.

Thermally Sprayed Coatings: Aluminum on Lead

An experimental program to determine the feasibility of thermally spraying aluminum on a lead substrate was initiated in support of the accelerator production of tritium (APT) Project for the U.S. Department of Energy. The program consisted of two distinct parts: (1) the characterization of the thermally sprayed coatings, including microhardness testing, effects of heating, and microstructure and porosity determinations, and (2) effects of mercury doping and heat treatments on the thermally sprayed composite. The project determined that aluminum could successfully be thermally sprayed onto the lead. The coatings had a dense microstructure, with a Vicker's Pyramid Hardness (VPH) of about 60, and a maximum porosity (found in strips on the samples) of 12%. Published by Elsevier Science Inc.

Water Chemistry Control for the Target/Blanket Region of the Accelerator Production of Tritium

High-energy particle interactions in the various components of the target/blanket region of the Accelerator Production of Tritium lead to heat generation and deposition. Heavy-water and light-water systems are used to cool the target/blanket system and associated equipment. Structural materials include Inconel alloy 718, aluminum-clad lead rods, aluminum tubes containing helium-3 and tritium gas, and stainless steel components. Proper coolant chemistry is required to maximize neutron production, minimize corrosion of components, and minimize activity buildup. Corrosion-related phenomena and development of coolant and moderator corrosion control for both power and defense fission reactors has been studied extensively over the past 50 years. Less is known, however, about cooling systems for accelerators where a variety of transient chemical species and spallation products may be formed. The following provides a discussion on the issues that need to be addressed for proper water chemistry control for the APT system.

Thermal analysis of the APT power coupler and similarities to superconducting magnet current leads

A detailed thermal analysis has been performed on the 210 kW, 700 MHz RF power coupler (PC) which transfers microwave energy from high power klystrons to the superconducting (SC) resonant cavities for the acceleration of protons. The work is part of the design for Accelerator Production of Tritium funded by the US Department of Energy. The PC is a co-axial design with the RF power transmitted in the annular region between two concentric cylinders. The PC provides a thermal connection from room temperature to superconducting niobium operating at 2.15 K. Heat transfer mechanisms considered are conduction, infra-red radiation, RF joule heating in normal and superconducting materials, and, forced and natural convection cooling. The objective of the thermal analysis is to minimize the required refrigeration power subject to manufacturability and reliability concerns. The problem is reminiscent of the optimization of superconducting magnet leads. The similarities and differences in the results between SC leads and PCs are discussed as well as the critical parameters in the PC optimization.