skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: The design of the DUPIC spent fuel bundle counter

Abstract

A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs.

Authors:
; ; ;
Publication Date:
Research Org.:
Los Alamos National Lab., NM (United States)
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
481874
Report Number(s):
LA-13239-MS
ON: DE97006374; TRN: 97:011209
DOE Contract Number:
W-7405-ENG-36
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: May 1997
Country of Publication:
United States
Language:
English
Subject:
44 INSTRUMENTATION, INCLUDING NUCLEAR AND PARTICLE DETECTORS; 05 NUCLEAR FUELS; HE-3 COUNTERS; PERFORMANCE TESTING; RESPONSE FUNCTIONS; SPENT FUEL ELEMENTS; CANDU TYPE REACTORS; ELECTRONIC EQUIPMENT; CURIUM 244; SAFEGUARDS

Citation Formats

Menlove, H.O., Rinard, P.M., Kroncke, K.E., and Lee, Y.G.. The design of the DUPIC spent fuel bundle counter. United States: N. p., 1997. Web. doi:10.2172/481874.
Copy to clipboard
Menlove, H.O., Rinard, P.M., Kroncke, K.E., & Lee, Y.G.. The design of the DUPIC spent fuel bundle counter. United States. doi:10.2172/481874.
Copy to clipboard
Menlove, H.O., Rinard, P.M., Kroncke, K.E., and Lee, Y.G.. Thu . "The design of the DUPIC spent fuel bundle counter". United States. doi:10.2172/481874. https://www.osti.gov/servlets/purl/481874.
Copy to clipboard
@article{osti_481874,
title = {The design of the DUPIC spent fuel bundle counter},
author = {Menlove, H.O. and Rinard, P.M. and Kroncke, K.E. and Lee, Y.G.},
abstractNote = {A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs.},
doi = {10.2172/481874},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu May 01 00:00:00 EDT 1997},
month = {Thu May 01 00:00:00 EDT 1997}
}
Copy to clipboard
Technical Report:
View Technical Report (1.91 MB)
DOI:  10.2172/481874
Save / Share:
Export Metadata
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.

  • This volume is one of a series of topical reports that summarize task activities and accomplishments by the General Electric Company for engineering and test support in the potential application of Morris Operation (MO) in 1980 and 1981 on a variety of fuel having different irradiation histories, different cooling times, and different reactor origins. The transfer rate tests performed at Morris Operation fell into two distinct categories: (1) the in-basin or qualitative tests and (2) the vessel or quantitative tests. The in-basin radionuclide transfer tests were performed in 1980 and were done in-situ; that is, the fuel bundles were notmore » moved, and measurements were made on the bundles while they were still in the storage basket. This method provided a means of qualifying those fuel bundles, or fuel categories, which were determined to be radiocesium or radiocobalt contributors. The vessel tests were performed in 1981 and utilized a specially fabricated underwater test vessel. The vessel tests allowed for quantitative measurements to be made on the radionuclide transfer rates. Two methods of running the vessel tests were used: (1) the closed-loop method and (2) the once-through method, which provided a means of determining the accuracy and repeatability of the measured transfer rates. Completion of this test program verified the actual operational experience at Morris Operation, that the release of either radiocobalt or radiocesium from spent fuel can be adequately controlled. The transfer rates measured for these radionuclides were easily controlled by the basin ion-exchange filtration system, and their release posed no problems to either the operating personnel or the general public.« less

  • ;
    This volume is one of a series of topical reports that summarize task activities and accomplishments for engineering and test support in the potential application of Morris Operation to the DOE Away-From-Reactor Spent Fuel Storage Program. Gamma exposure rates near spent fuel were measured in both air and water and are reported in this volume. These rates are compared with exposure rates calculated from parameters routinely supplied by the utility (i.e. initial enrichment, average power density, burnup and decay time). The ORIGEN and QAD-F computer codes were used for the calculation. For measurements, an apparatus was made to hold amore » fuel bundle and an ion chamber underwater in any of several fixed geometries. A diving bell sized to fit ovr both the fuel and detector was used for in-air measurements, because the water around the fuel could be displaced downward with compressed air. Maximum exposure rates varied from 36,000 R/h next to a 3.5 year old PWR bundle with a 30,000 MWD/MT burnup, to 40 R/h for a 10.5 year old BWR bundle with a burnup by only 180 MWD/MT. Exposure rates in air were only slightly higher than those in water for positions close to the bundles. Farther away where the water shielding was greater the difference was also greater. Reasonable agreement was seen between calculated and measured values (i.e. within +- 50%) for cases where scattering was of secondary importance. The quality of the agreement reflects the detail considered in setting up the geometry for the computer runs.« less

  • ;
    Thermal output measurements were made on fourteen spent fuel bundles at Morris Operation in 1981, and the results are reported in this volume. These thermal output rates are compared with predicted thermal values. The thermal measurements were made in a specially fabricated test vessel which was located underwater in the Morris Operation storage basin. A total of twenty-four determinations were made on the fourteen fuel bundles which were selected for study. The fuel bundles studied had burnup values ranging from 26,400 to 40,500 MWD/MTU, and cooling times ranging from 4.5 to 8.25 years. Comparisons made between the experimentally measured thermalmore » values and the mathematically predicted values indicated that the proposed ANS-5 standard, revised 1973, came closest to the measured values for the relatively short-cooled fuel (< 2000 days) and that the ORIGEN code came closest for the longer-cooled fuel (> 2000 days). Additional improvement of the ORIGEN code predicted values was demonstrated when the calculations were based upon the actual irradiation history, rather than on the assumption of a constant specific power irradiation history. The total error associated with these thermal measurements was +- 1% for the range of thermal outputs observed (360 to 940 watts). The accuracy of the thermal values was demonstrated by making the thermal determinations via two independent methods.« less

  • Results of an adiabatic, single bundle test program to determine the subcooled counter-current flow limiting (CCFL) characteristics of the upper region of a BWR fuel bundle are presented. Data are reported for three methods of inlet water injection and three methods of steam injection. The results show that the CCFL characteristics are affected by the methods of both steam and water injection. It is also shown that CCFL with subcooled liquid injection is characterized by the steam condensing capability of the inlet water.

  • ; ;
    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuelmore » Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.« less