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in subjects: "Hastelloy X"

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Reactor safety research programs. Quarterly progress report, January 1--March 31, 1977

The projects reported each quarter are the following: Gas Reactor Safety Evaluation, THOR Code Development, SSC Code Development, LMFBR and LWR Safety Experiments, Fast Reactor Safety Code Validation, Technical Coordination of Structural Integrity, and Fast Reactor Safety Reliability Assessment.

Date: May 1, 1977

Creator: Romano, A. J. (comp.)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Ceramic coatings for components exposed to coal-gas environments: a review. [64 references]

The corrosive and erosive environments at high temperatures and pressures in coal gasifiers impose severe requirements on the alloys of fabrication. A concise review of the application of ceramic coatings to resist coal-gas environments has been conducted. The purpose of this review is to explore suitable ceramic or cermet materials that may resist or retard the degradation of metal components and to summarize the state of the art of various methods of producing such coatings.

Date: December 1, 1976

Creator: Swaroop, R

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Creep-rupture properties and corrosion behavior of 2 1/4 Cr-1 Mo steel and Hastelloy X alloys in simulated HTGR environment-interim report

Hastelloy X and 2/sup 1///sub 4/ Cr-1 Mo steel are being considered as structural alloys for components of a High-Temperature Gas-Cooled Reactor (HTGR) system. Among other mechanical properties, the creep behavior of these materials in HTGR primary coolant helium must be established to form part of the design criteria. This report describes the simulated HTGR-helium environmental creep facilities, summarizes preliminary creep properties of 2/sup 1///sub 4/ Cr-1 Mo steel and Hastelloy X generate… more

Date: November 1, 1977

Creator: Lystrup, A. S.; Rittenhouse, P. L. & DiStefano, J. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Application of Hastelloy X in gas-cooled reactor systems

Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with co… more

Date: October 1, 1976

Creator: Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W. R.; Strizak, J. P.; Lystrup, A. & DiStefano, J. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Selection of non-adsorbing alkali components

This project consists of three phases of laboratory experimental study. In phase I (screening), eight candidate materials, 304SS (serves as a base material for comparison), Hastelloy C-276, Hastelloy X, Haynes No. 188, Allonized 304SS, Pt-coated 304SS, and ceramic-coated 304SS, will be subjected to atmospheric TGA study under the simulated PFBC (oxidizing) environment with and without alkali vapor doping. Each candidate material will be evaluated for its resistance toward alkali-vapor capture. … more

Date: January 1, 1992

Creator: Lee, S. H. D.; Natesan, K. & Swift, W. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Production test IP-376-D, Supplement B Irradiation of MGCR-HDR-3 Test Element

The objective of this supplement to PT-IP-376-D, Irradiation of MGCR-HDR-3 Test Element is to authorize 1000 hours of operation at a maximum test specimen surface temperature of 1700 F. The original production test authorized a test duration of four months at a maximum specimen surface temperature of 1500 F; supplement A authorized extension of the test duration to ten months. The desired increase in surface temperature is requested to demonstrate the general feasibility of operation of the fue… more

Date: July 11, 1961

Creator: Baars, R. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Selection of non-adsorbing alkali components

Date: November 1, 1992

Creator: Lee, S. H. D.; Natesan, K. & Swift, W. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effects of irradiation and thermal aging upon fatigue-crack growth behavior of reactor pressure boundary materials. [Neutrons]

Two processes that have the potential to produce degradation in the properties of pressure boundary materials are neutron irradiation and long-time thermal aging. This paper uses linear-elastic fracture mechanics techniques to assess the effect of these two processes upon the fatigue-crack growth behavior of a number of alloys commonly employed in reactor pressure boundaries. The materials evaluated include ferritic steels, austenitic stainless steels, and nickel-base alloys typical of those em… more

Date: October 1, 1978

Creator: James, L. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Study of tertiary creep instability in several elevated-temperature structural materials

Data for a number of common elevated temperature structural materials have been analyzed to yield mathematical predictions for the time and strain to tertiary creep at various rupture lives and temperatures. Materials examined include types 304 and 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, alloy 718, Hastelloy alloy X, and ERNiCr--3 weld metal. Data were typically examined over a range of creep temperatures for rupture lives ranging from less than 100 to greater than 10,000 hours. W… more

Date: January 1, 1978

Creator: Booker, M. K. & Sikka, V. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

Date: January 1, 1980

Creator: Romano, A.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Brayton Isotope Power System (BIPS) superalloy ground demonstration system (GDS-S) detail design review

The material presented at the GDS-S design review meeting held at Phoenix, Arizona on September 7 and 8, 1977, is reported. This design review was specifically scoped to examine the Hastelloy-X Heat Source Heat Exchanger (HSHX) and the Hastelloy-X bellows. These are new components required to upgrade the Workhorse Loop (WHL) to the GDS configuration. Additional topics covered in support of those items were: reliability; materials; and the WHL-to-GDS conversion sequence.

Date: September 12, 1977

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechan… more

Date: November 24, 1978

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differe… more

Date: January 1, 1981

Creator: Swindeman, R.W. & Brinkman, C.R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Evaluation of creep and relaxation data for hastelloy alloy x sheet

Hastelloy alloy X has been a successful high-temperature structural material for more than two decades. Recently, Hastelloy alloy X sheet has been selected as a prime structural material for the proposed Brayton Isotope Power System (BIPS). The material also sees extensive application in the High-Temperature Gas-Cooled Reactor (HTGR). Design of these systems requires a detailed consideration of the high-temperature creep properties of this material. Therefore, available creep, creep-rupture, an… more

Date: February 1, 1979

Creator: Booker, M. K.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effect of fission product interactions on the corrosion and mechanical properties of HTGR alloys. [HTGR]

Preliminary experiments have been carried out to determine how fission product interactions may influence the mechanical integrity of reference HTGR structural metals. In this work Type 304 stainless steel, Incoloy 800 and Hastelloy X were heated to 550 to 650/sup 0/C in the presence of CsI. It was found that no corrosion of the alloys occurred unless air or oxygen was also present. A mechanism for the observed behavior is proposed. A description is also given of some long term exposures of HTG… more

Date: January 1, 1978

Creator: Aronson, S.; Chow, J.G.Y.; Soo, P. & Friedlander, M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Correlation of high cycle and low cycle fatigue data for some HTGR structural metals

An analytical procedure has been evaluated to determine whether low and high cycle fatigue testing techniques may be correlated in the 10/sup 5/ cycle region where the data overlap. The procedure, which is based on the use of cyclic stress-strain curves to convert high cycle fatigue stresses to equivalent strains, is shown to be acceptable for Incoloy 800H, Hastelloy X, Type 304 stainless steel and 2 1/4 Cr--1Mo steel in the range of temperature for which data are available.

Date: January 1, 1978

Creator: Soo, P. & Chow, J.G.Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Relationship between carburization and zero-applied-stress creep dilation in Alloy 800H and Hastelloy X. [HTGR]

Typical HTGR candidate alloys can carburize when exposed to simulated service environments. The carbon concentration gradients so formed give rise to internal stresses which could cause dilation. Studies performed with Hastelloy X and Alloy 800H showed that dilations of up to almost 1% can occur at 1000/sup 0/C when carbon pickup is high. Dilation was normally observed only when the carbon increase was >1000 ..mu..g/cm/sup 2/ and ceased when the diffusing carbon reached the center of the specim… more

Date: January 1, 1981

Creator: Inouye, H. & Rittenhouse, P.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

High-temperature low-cycle fatigue and tensile properties of Hastelloy X and alloy 617 in air and HTGR-helium

Results of strain controlled fatigue and tensile tests are presented for two nickel base solution hardened alloys which are reference structural alloys for use in several high temperature gas cooled reactor concepts. These alloys, Hastelloy X Inconel 617, were tested at temperatures ranging from room temperature to 871/sup 0/C in air and impure helium. Materials were tested in the solution annealed as well as in the pre-aged condition where aging consisted of isothermal exposure at one of sever… more

Date: January 1, 1981

Creator: Strizak, J.P.; Brinkman, C.R. & Rittenhouse, P.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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TRU waste cyclone drum incinerator and treatment system: January--March 1978

The cyclone incinerator was operated throughout the past quarter, generating additional data on system characteristics, equipment life expectancies, and by-product generation. Several changes in the incinerator system are in various stages of completion. The lid assembly, secondary chamber, and expansion unit for the new exhaust equipment are nearly ready for installation. A new heat exchanger has been installed in the scrubber system. An ash handling system has been designed for possible futur… more

Date: May 5, 1978

Creator: Klingler, L.M.; Batchelder, D.M. & Lewis, E.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thulium oxide fuel characterization study: Part 1, Materials properties measurements. [Tm/sub 2/O/sub 3/-Yb/sub 2/O/sub 3/; thulium-170]

A feasibility study was performed on encapsulated thulium-170 as an isotopic fuel for operation at temperatures to 1500/sup 0/C for 180 days. Effects of various combinations of fueled capsule design parameters were evaluated and compard to experimental data. A computer program was developed to predict dose rates through various thicknesses of aluminum, stainless steel, lead, tungsten and depleted uranium absorbers using thermoluminescent dosimetry techniques for experimental corroboration. A pr… more

Date: August 1, 1970

Creator: Nelson, C.A.; Anderson, R.W.; Fink, C.R.; Tse, A. & Fretague, W.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

HTGR structural-materials efforts in the US

The status of ongoing structural materials programs being conducted in the US to support development and deployment of the high-temperature gas-cooled reactor (HTGR) is described. While the total US program includes work in support of all variants of this reactor system, the emphasis of this paper is on the work aimed at support of the steam cycle/cogeneration (SC/C) version of the HTGR. Work described includes activities to develop design and performance prediction data on metals, ceramics, an… more

Date: July 1, 1982

Creator: Rittenhouse, P.L. & Roberts, D.I.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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HTGR Generic Technology Program: materials technology reactor; operating experience; medium-enriched-uranium fuel development. Quarterly progress report for the period ending July 31, 1978

The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.

Date: August 1, 1978

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Gas-Cooled Reactor Programs. High-Temperature Gas-Cooled Reactor Base-Technology Program. Annual progress report for period ending December 31, 1977

Progress in HTGR studies is reported in the following areas: fission product technology and coolant impurity effects, fueled graphite development, PCRV development, structural materials, characterization and standardization of graphite, and evaluation of the pebble-bed type HTR.

Date: July 1, 1978

Creator: Homan, F. J. & Kasten, P. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Weldability evaluations and weldment properties of Hastelloy X

Studies of weldability and weldment properties were conducted on commerical heats of Hastelloy X. Weldment preparation was done using several combinations of welding techniques and filler metals. Evaluation methods employed included hot cracking susceptibility and tensile and creep properties measured both before and after aging at 593 to 871/sup 0/C for up to 10,000 h.

Date: January 1, 1981

Creator: King, J. F.; McCoy, H. E. & Rittenhouse, P. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department