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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

FLiNaK Compatability Studies with Inconel 600 and Silicon Carbide

A small liquid fluoride salt test apparatus has been constructed and testing conducted to examine the compatibility of SiC, Inconel 600, and a spiral wound gasket material in FLiNaK salt. These tests were conducted to test materials and sealing systems that would be used in a FLiNaK salt test loop. Three months of testing at 700oC was used to assure that these materials and seals would be acceptable operating under expected test loop conditions. The SiC specimens showed little or no change over the test period while the spiral wound gasket material showed no degradation, except for the possibility of salt seeping into the outermost spirals of the gasket. The Inconel 600 specimens showed regions of voiding which penetrated the specimen surface to about 250 m in depth. Analysis indicated that the salt had leached chrome from the Inconel surface as was expected for this material. Because the test loop will have a limited working lifetime, it was concluded that these materials would be satisfactory for loop construction.

42 ENGINEERING↗
DOE OSTI2019

The Effect of Weld Metal Composition on Microstructural and Mechanical Properties of Dissimilar Welds Between Monel 400 and Inconel 600

The current work was carried out to characterize welding of Inconel 600 superalloy and Monel 400. In the present study, gas tungsten arc welding (GTAW) with four types of filler metals (ENiCu, ENiCu-7, ENiCr-3 and ENiCrFe-3) was utilized. The filler metals were categorized into two groups based on the presence of Cr and Nb in their chemical composition. This paper describes the selection and application of welding consumables in manufacturing dissimilar metal combinations. It was found that (Nb/Cr) ratio and Fe content can affect on the size and amount of carbides. During solidification of ENiCr-3 and ENiCrFe-3 filler metals, Nb leaves dendritic cores and is rejected to inter-dendritic regions because of its high tendency for segregation. However, No brittle Laves phase was detected in the welds. The interface between weld metal and base metal and gradual changes in alloying elements were also investigated by scanning electron microscopy. EDX analysis was also applied for different samples. Microhardness measurements and tensile test were performed on the joints. The average hardness value in the weld metal obtained from ENiCrFe-3 is higher than the other welds. ENiCu-7 weld exhibited the highest tensile strength of 540 MPa and percent elongation of 53% in contrast to ENiCrFe-3 weld with the lowest tensile strength (518 MPa) and percent elongation (39%).

36 MATERIALS SCIENCE↗
DOE OSTI2022

High-Temperature RF Transmission Loss Characteristics of Platinum–Inconel 600 and Platinum–304 Steel Interconnects

Radio frequency (RF) transmission losses that occur at electronic interconnects is currently an important limitation of wireless sensors and other electronic device packages that must operate in high-temperature environments. Herein, in-situ RF transmission losses were measured for two types of interconnects: platinum (Pt) wire bonded to an Inconel 600 coaxial cable; and Pt-wire bonded to a block of 304 steel. Heating the Pt wire/Inconel interconnects at 800 °C for up to 200 hours were found to be very stable with no measurable transmission losses. In contrast, Pt wire/304 steel block interconnects examined over similar test periods at 800 °C exhibited gradual increases in transmission losses leveling off at 10, 38 and 85 mdB at 400 MHz, 700 MHz, and 1 GHz, respectively. Increasing the test temperature to 900 °C caused further elevated transmission losses of 30, 70 and 108 mdB at the same reported frequencies. By comparison, temperature cycling of the Pt wire/304 steel interconnects between room temperature and 900 °C was found to cause more significant degradation. In particular, four thermal cycles from room temperature to 900 °C resulted in losses of 75, 271, and 482 mdB at the same reported frequencies. Parabolic oxidation relative to the RF skin depth in the 304 steel block at the tested frequencies are discussed as possible reasons for the transmission losses during the isothermal soaks, whereas interconnect degradation due to differences in thermal expansion between Pt and 304 steel are primarily responsible for transmission losses caused by high temperature cycling.

304 steel↗
DOE OSTI2022

Analysis of inconel 600 oxidized under loss-of-coolant accident conditions: A multi-modal approach

Understanding the performance of cladding materials during a loss-of-coolant accident scenario is crucial for developing next-generation accident tolerant nuclear fuel rod claddings. We describe a multi-modal approach combining synchrotron-based diffraction, and Raman spectroscopy with nano-scale electron microscopy techniques for investigating oxidation of Inconel 600 (A600) in steam and air environments at 1200 °C for 2 h. We report that A600 exposed to steam develops a Cr 2 O 3 -enriched surface layer while a mixed phase oxide layer containing NiFe 2 O 4 , Cr 2 O 3 , Fe 3 O 4 and Fe (3–x) CrxO 4 is formed in air. Mechanism of oxidation for A600 in air and steam environments is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Corrosion sensitivity of nickel-based Alloy Inconel 600 in pressurized water reactor water chemistry: Can KOH replace LiOH?

Lithium hydroxide (LiOH) has been used to balance water acidity against boric acid moderators in the primary water of Western pressurized water reactor (PWR) designs for decades. However, the demand and the cost of lithium-7 has grown significantly since 2015. Potassium hydroxide (KOH) has been identified as a suitable and cost-effective substitute for LiOH and has been used successfully in Russian PWR designs for more than 40 years. However, it is important to know if alloys and water conditions used in Western PWR’s (Alloy 600) are similarly compatible with KOH additions. This work is focused on the aqueous corrosion behaviors of Alloy 600 with LiOH versus KOH additions at normal operating concentrations and crevice water chemistry at simulated PWR primary water conditions. TEM was used to characterize the formed oxide through diffraction analysis; SIMS was used to probe the cation ingress into the material; Atom probe tomography was used to determine the 3D elemental distribution within the oxide/metal structures; finally electrochemical impedance spectroscopy was used to discuss the structure and corrosion resistance of the oxide films. These advanced characterization techniques complement the weight measurments which showed lower mass gain in KOH than LiOH crevice water due to less oxide formation. Overall, the use of KOH as a potential alternative to LiOH in PWR is discussed.

Alloy 600↗
DOE OSTI2014

Microstructural Development and Mechanical Properties for Reactive Air Brazing of ZTA to Ni Alloys using Ag-CuO Braze Alloys

Reactive air brazing (RAB) is a potential joining technique to join metal alloys to ceramics for a variety of applications. In the current study, nickel (Ni) alloys were heat treated to form an oxide layer prior to RAB joining to zirconia toughened alumina (ZTA). The Ni alloys evaluated were Nicrofer 6025 HT, Inconel 600, Inconel 693, Haynes 214 and Inconel 601. The ZTA studied had compositions of 0 to 15 wt% zirconia and 0 to 14 wt% glass. Four point-bend tests were performed to evaluate the joint strength of ZTA/ZTA and ZTA/nickel alloys brazed with Ag-2wt% CuO braze alloys. It was determined that the joint strength is not a function of the ZTA composition, but that the strength is a strong function of the chemistry and microstructure of the oxide layer formed on the nickel alloy. It was determined that an increase in the aluminum content of the Ni alloy resulted in an increase of the thickness of alumina in the oxide layer and was directly proportional to the bond strength with the exception of Inconel 601 which exhibited relatively high joint strengths even though it had a relatively low aluminum content.

Ag-CuO↗
DOE OSTI2019

Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel

Steady and transient (with a heating rate of 685?/s) internal-flow CHF (Critical Heat Flux) experiments were conducted under atmospheric pressure at a fixed inlet temperature (40? or 60?) and mass flow (300kg/m 2 s) on Fe-13Cr-6Al alloy, Inconel 600 and Zircaloy-4 tube samples. Multiple experiments were repeated on the same specimen to investigate the effect of surface characteristic changes (i.e., roughness, wettability, and oxide scale morphology) on the occurrence of CHF. Despite notable changes of wettability, roughness, and oxide layer characteristics on samples that had already been subjected to CHF, measured flow CHF remained unchanged throughout repeated experiments for tested materials. This demonstrates that the surface effects on flow CHF are limited in the test conditions. In the steady-state flow boiling condition, Fe-13Cr-6Al alloy demonstrated a 22% and 14% increase in CHF compared to Zircaloy-4 and Inconel 600, respectively. Compared to the 2006 Groeneveld CHF lookup table, Fe-13Cr-6Al alloy gives a 13% increase in the tested flow boiling condition. Material properties are considered primarily responsible for the observed CHF differences among the tested materials. The surface thermal economy parameter $\frac{(pc_p)^{3/2}}{\sqrt{k}}$ is proposed as an explanation for the observed CHF differences; this parameter is related to material’s ability to avoid an irreversible dry spot formation. The apparent disagreement of Zircaloy-4 CHF with both the look up table predictions and Inconel 600 shows the limitation of departure of nuclate boiling (DNB) evaluations that do not consider cladding materials. The transient Fe-13Cr-6Al CHF is 39% and 23% higher than the lookup table prediction and the steady-state condition experimental result, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of Cambridge have been investigated. The rationale for the superior performance of the type N using a customized sheath developed at the University of Cambridge is explained in comparison with the behavior of conventional type N Inconel 600 sheathed thermocouples. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Optical Fier Based System for Multiple Thermophysical Properties for Glove Box, Hot Cell and In-Pile Application

Thermal diffusivity of materials is of interest in nuclear applications at temperatures in excess of 2000°C. Commercial laser flash apparatus (LFA) that heats samples with a furnace typically do not reach these elevated temperatures nor are they easily adapted to a glove-box or hot cell environment. In this research, we performed work on an experimental technique using single laser surface heating, i.e. heating the disk sample only at its front surface with the continuous wave (CW) laser, to allow measurement of thermal diffusivity at very high temperatures within a small chamber. Thermal diffusivity is measured using a separate pulsed laser on the front side and IR detector on the rear side. The new way of heating provides easy operation in comparison to other heating methods. The measurement of sample reference temperature is needed for the measured thermal diffusivity. A theoretical model was developed to describe transient heat transfer across the sample due to the laser pulse, starting from the steady state temperature of the sample heated by the CW laser. The experimental setup was established with a 500W CW laser and maximum 50 Joule pulse laser irradiated at the front surface of the sample. The induced temperature rise at the rear surface, along with the steady-state temperature at the front surface, was recorded for the determination of thermal diffusivity and the sample temperature. Three samples were tested in vacuum over a wide temperature range of 500°C to 2100°C, including graphite, Inconel 600 and tungsten. The latter two samples were coated with sprayed graphite on their front surfaces in order to achieve surface absorption/emission needs, i.e. high absorptivity of the front surface against relatively low emissivity of the rear surface. Thermal diffusivity of graphite determined by our system are within a 5% difference of the commercial LFA data at temperatures below 1300°C and agree well with its trend at higher temperatures. Good agreement would also exist for Inconel 600 and tungsten. Despite large uncertainty of measuringthe sample temperature, the uncertainties of thermal diffusivity are less than 6% for all samples at elevated temperatures. The results indicate that single laser surface heating could be convenient and practical for the application of the LFA measurements without extra uncertainty, as temperature dependence of thermal diffusivity is usually negligible in the sample. Moreover, it is concluded that unequal surface treatment, i.e., high absorption on the front side and low emission on the rear side, greatly improves the measurement in serval aspects: less power requirement of the CW laser, less uncertainty of measured thermal diffusivity, and more uniform temperature distribution in the sample. The result of this research can be used as a general guideline for the design of this type of measurement system for nuclear applications. It can also be used directly to design and build a system similar to the one implemented in this project.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Experimental investigations of Critical Heat Flux re-occurrence on post-CHF surfaces

Under the extreme environments of light water reactors, the cladding materials demonstrate evolutionary transformations in terms of material compositions, surface morphologies, and thermal-physical properties. This study investigates how the temperature overshooting of Critical Heat Flux (CHF) affects the re-occurrence of CHF on the post-CHF surfaces. Experimental results reveal that the CHF difference between as-received and post-CHF surfaces can be significant, which depends on materials. CHF is enhanced on post-CHF Stainless Steel (SS 316) specimens while it deteriorates on post-CHF Inconel-600 specimens. In comparison with the as-received surfaces, the post-CHF surface of SS 316 is more hydrophobic, while Inconel 600 is more hydrophilic. The scanning electron microscope micrographs of post-CHF surfaces show that surface morphologies that are structured by oxide crystals are formed on SS 316 samples but not on Inconel 600 samples. The physical rationale behind the CHF difference could be boiling heat transfer conjugated with thermal-physical proprieties of investigated surface materials and surface morphologies features. In conclusion, the results imply that it is indispensable to bookkeep the evolutionary changes of cladding surface materials, especially for cladding candidates of Accident Tolerant Fuel since operation histories can leave footprints on cladding surfaces, thus resulting in CHF re-occurrence differences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Hot Corrosion Behavior of Ni20Cr Alloy in NaVO{sub 3} Molten Salt

The hot corrosion behavior of Ni20Cr alloy was evaluated in NaVO{sub 3} at temperatures of up to 700 °C and compared to Inconel 600. The performance of both the alloys was evaluated using electrochemical measurements, such as potentiodynamic polarization, linear polarization resistance, and electrochemical impedance spectroscopy measurements, as well as a cyclic corrosion test. The results show that the Ni20Cr alloy has better corrosion performance than the Inconel 600. The better performance of the Ni20Cr alloy was due to its ability to develop a stable protector oxide (Cr{sub 2}O{sub 3}) on its surface.

36 MATERIALS SCIENCE↗
DOE OSTI2015

NEET-AMM Final Technical Report on Laser Direct Manufacturing (LDM) for Nuclear Power Components

Final technical report summarizes the program progress and technical accomplishments of the Laser Direct Manufacturing (LDM) for Nuclear Power Components project. A series of experiments varying build process parameters (scan speed and laser power) were conducted at the outset to establish the optimal build conditions for each of the alloys. Fabrication was completed in collaboration with Quad City Manufacturing Laboratory (QCML). The density of all sample specimens was measured and compared to literature values. Optimal build process conditions giving fabricated part densities close to literature values were chosen for making mechanical test coupons. Test coupons whose principal axis is on the x-y plane (perpendicular to build direction) and on the z plane (parallel to build direction) were built and tested as part of the experimental build matrix to understand the impact of the anisotropic nature of the process.. Investigations are described 316L SS, Inconel 600, 718 and 800 and oxide dispersion strengthed 316L SS (Yttria) alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

High strength kiloampere Bi 2 Sr 2 CaCu 2 O x cables for high-field magnet applications

Multifilamentary Ag-sheathed Bi 2 Sr 2 CaCu 2 O x (Bi-2212) wire can carry sufficient critical current density J c for the development of powerful superconducting magnets. But, the range of its applications is limited by the low mechanical strength of the Ag/Bi-2212 strand. A potential solution is to cable Ag/Bi-2212 wire with high-strength materials that are compatible with the Bi-2212 heat treatment in an oxygen atmosphere. Past attempts have not always been successful, because the high-strength materials reacted with Bi-2212 wires, significantly reducing their J c . We examined the nature of reactions occurring when Ag/Bi-2212 wires are heat-treated in direct contact with several commonly used high-strength alloys and a new Fe-Cr-Al alloy. INCONEL X750 and INCONEL 600 resulted in significant J c loss, whereas Ni80-Cr caused little or no J c loss; however, all of them formed chromium oxide that subsequently reacted with silver, creating cracks in the silver sheath. We found that Fe-Cr-Al did not show significant reactions with Ag/Bi-2212 strands. Scanning electron microscopy (SEM) and energy dispersive x-ray (EDS) examinations revealed that the Fe-Cr-Al alloy benefits from the formation of a uniform, crack-free, continuous alumina layer on its surface that does not react with Ag and that helps minimize the Cu loss found with INCONEL X750 and INCONEL 600. We then fabricated prototype 6-around-1 cables with six Bi-2212 strands twisted and transposed around an Fe-Cr-Al alloy core coated with TiO 2 . After standard 1 bar melt processing, the cable retained 100% of the total current-carrying capability of its strands, and, after a 10 bar overpressure processing, the cable reached a total current of 1025 A at 4.2 K and 10 T. Tensile tests showed that Fe-Cr-Al becomes brittle after being cooled to 4.2 K, whereas INCONEL X750 remains ductile and retains a modulus of 183 GPa. Finally. we proposed new cable designs that take advantage of the chemical compatibility of Fe-Cr-Al and high strength of INCONEL X750 for various high-field magnet applications.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties

In this paper we analyze the differences between previous transient critical heat flux (CHF) experiments using iron-chromium-aluminum (FeCrAl), Inconel 600, and stainless steel 316 (SS316) alloy test sections and best-estimate modeling results from widely-used nuclear engineering systems and subchannel analysis tools. FeCrAl is an Accident Tolerant Fuel (ATF) candidate cladding material. The thermal hydraulic performance and safety characteristics of FeCrAl are being evaluated to determine viability as a cladding material in Light Water Reactors (LWRs). In this study, the results of the CHF experiments conducted at atmospheric pressure and fixed inlet coolant temperature and mass flux are compared to models built in the fifth version of the Reactor Excursion and Leak Analysis Program (RELAP5-3D) and CTF, the modernized version of COBRA-TF developed by the Consortium for Advanced Simulation of LWRs (CASL). Results from RELAP5-3D and CTF showed differences from the experiments and from each other in predicting CHF. In the Inconel 600 case, both computational tools overpredicted CHF, which led to an underprediction in the tube outer surface temperature. In the SS316 and FeCrAl cases, CHF was underpredicted by the codes, leading to an overprediction of the tube outer surface temperature. To understand the discrepancies in CHF and post-CHF predictions, studies were performed using RELAP5-3D and RAVEN to determine the sensitivity of CHF and peak test section temperature, an analog to peak cladding temperature (PCT), to heat transfer coefficients, a CHF multiplier, and uncertainties in the thermal conductivity and volumetric heat capacity. We found that CHF depends most strongly on the CHF multiplier and thermophysical properties. A combination of these factors that produced the best match to the experiment based on CHF, PCT, and the total energy deposited into the tube was determined. The best match parameters were able to provide best-estimate predictions of the CHF and integral heat flux, but were still conservative when predicting the PCT. The best match set of parameters developed in this paper are intended only as a demonstration of an approach that could be applied in the future with a larger set of experiments to produce more accurate models of CHF and post-CHF behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Microstructure, Processing, Performance Relationships for High Temperature Coatings

This work evaluates the suitability of iron aluminide coatings for use in high temperature fossil fuel combustion environments, such as boiler applications. The coatings are applied using High Velocity Oxy-Fuel (HVOF) thermal spray techniques. Iron aluminide coatings, with the nominal composition of Fe3Al, were applied to various high temperature structural materials (316 Stainless Steel, 9Cr-1Mo steel and Inconel 600) that typically lack inherent resistance to environmental degradation found in fossil fuel combustion atmospheres. Coating/substrate combinations were subjected to thermal cycling to evaluate the effect of HVOF parameters, coating thickness, substrate material and substrate surface roughness on the resistance to coating delamination and cracking. It was found that substrate surface roughness had a profound influence on the performance of a given substrate/coating system and that surface preparation techniques will need to be tailored to the specific substrate material. Also, higher particle velocity during HVOF thermal spray deposition of the iron aluminide coatings tended to result in better-performing coating/substrate systems with less delamination at the coating/substrate interface. Some combinations of HVOF parameters, coating thickness and substrate materials were found to perform extremely well even at temperatures up to 900oC. However, in some cases, substantial reactions at the interface were observed.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2011

Shielding experiments by the JASMIN Collaboration at Fermilab (II) - radioactivity measurement induced by secondary particles from the anti-proton production target

The JASMIN Collaboration has performed an experiment to conduct measurements of nuclear reaction rates around the anti-proton production (Pbar) target at the Fermi National Accelerator Laboratory (FNAL). At the Pbar target station, the target, consisting of an Inconel 600 cylinder, was irradiated by a 120 GeV/c proton beam from the FNAL Main Injector. The beam intensity was 3.6 x 10 12 protons per second. The samples of Al, Nb, Cu, and Au were placed near the target to investigate the spatial and energy distribution of secondary particles emitted from it. After irradiation, the induced activities of the samples were measured by studying their gamma ray spectra using HPGe detectors. The production rates of 30 nuclides induced in Al, Nb, Cu, Au samples were obtained. These rates increase for samples placed in a forward (small angle) position relative to the target. The angular dependence of these reaction rates becomes larger for increasing threshold energy. These experimental results are compared with Monte Carlo calculations. The calculated results generally agree with the experimental results to within a factor of 2 to 3.

36 MATERIALS SCIENCE↗