40-MW(e) Prototype High-Temperature Gas-Cooled Reactor Research and Development Program. Quarterly Progress Report for the Period Ending June 30, 1962

PDF Version Also Available for Download.

Description

Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 ... continued below

Physical Description

Pages: 61

Creation Information

Creator: Unknown. October 31, 1963.

Context

This report is part of the collection entitled: Office of Scientific & Technical Information Technical Reports and was provided by UNT Libraries Government Documents Department to Digital Library, a digital repository hosted by the UNT Libraries. More information about this report can be viewed below.

Who

People and organizations associated with either the creation of this report or its content.

Creator

  • We've been unable to identify the creator(s) of this report.

Publisher

Provided By

UNT Libraries Government Documents Department

Serving as both a federal and a state depository library, the UNT Libraries Government Documents Department maintains millions of items in a variety of formats. The department is a member of the FDLP Content Partnerships Program and an Affiliated Archive of the National Archives.

Contact Us

What

Descriptive information to help identify this report. Follow the links below to find similar items on the Digital Library.

Description

Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133 Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two basic assumptions made in the one-dimensional burnup code FEVER was investigated. As a result of extensive lifetime studies and power-distribution and temperaturecoefficient calculations, the initial fuel loading for the Peach Bottom core was specified. A series of control-rodworth calculations and a recalculation of the postulated rod-fall accident were made for this loading. The test of the prototype control rod and drive was satisfactorily coNonempleted during the quarter. During the course of the test the drive completed 590,641 starts and stops, 5,756 scrams, and more than 2.6 million inches of random regulating motion in helium at reactor temperatures. These totals far exceed the expected life requirements of the system. Preparations are being made for testing the prototype emergency shutdown rod and drive. The apparatus for the barium permeation experiment with a full-diameter sleeve was completed, and preliminary calibration runs were started. Following these runs, the system will be operated until an equilibrium distribution of barium is reached. At that time, a series of corings will be made on all of the compacts and the sleeve to evaluate the overall barium and strontium distribution. Other experiments on barium behavior, including permeation experiments with reducedscale fuel elements and exVeriments on the vaporization, sorption, and diffusion of barium, were continued, and the data are being analyzed. Measurements were made to compare the room-temperature back diffusion of argon, krypton, and xenon through a sample of sleeve graphite against a helium pressure difference. The results show that the difference between the effective back-diffusion coefficients of krypton and xenon seems to increase with increasing helium pressure difference across the sleeve. The argon and krypton back-diffusion data at an average pressure of 3 atm are essentially the same, A FORTRAN code was written to recalculate the retention of neutron poison material and fission products in the core as well as their condensation on and revaporization from the upper reflector following a complete loss-of-coolant-circulation accident. (auth)

Physical Description

Pages: 61

Source

  • Other Information: Orig. Receipt Date: 31-DEC-63

Language

Item Type

Identifier

Unique identifying numbers for this report in the Digital Library or other systems.

  • Report No.: GA-3396
  • Grant Number: AT(04-3)-314
  • DOI: 10.2172/4684878 | External Link
  • Office of Scientific & Technical Information Report Number: 4684878
  • Archival Resource Key: ark:/67531/metadc1028237

Collections

This report is part of the following collection of related materials.

Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

What responsibilities do I have when using this report?

When

Dates and time periods associated with this report.

Creation Date

  • October 31, 1963

Added to The UNT Digital Library

  • Oct. 18, 2017, 7:39 a.m.

Description Last Updated

  • Dec. 20, 2017, 1:34 p.m.

Usage Statistics

When was this report last used?

Congratulations! It looks like you are the first person to view this item online.

Interact With This Report

Here are some suggestions for what to do next.

Start Reading

PDF Version Also Available for Download.

International Image Interoperability Framework

IIF Logo

We support the IIIF Presentation API

40-MW(e) Prototype High-Temperature Gas-Cooled Reactor Research and Development Program. Quarterly Progress Report for the Period Ending June 30, 1962, report, October 31, 1963; San Diego, California. (digital.library.unt.edu/ark:/67531/metadc1028237/: accessed November 19, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.