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Fuel performance annual report for 1989

This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.

Date: June 1, 1992

Creator: Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)) & Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more

Date: January 1, 1982

Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Long-term leaching of irradiated spent fuel

Spent Light Water Reactor (LWR) fuel with burnups of 9, 28 and 54 MWd/kg U were leach tested at 25/sup 0/C in deionized water in a Paige apparatus. No discernible differences in leach rates were observed due to burnup. Additionally, the 28 MWd/kg U fuel was IAEA leach tested in five different leachants using the IAEA method. Deionized water gave the highest leach rates and a calcium chloride solution gave the lowest leach rates. An accelerated leaching period was observed during the Paige leach… more

Date: January 1, 1979

Creator: Katayama, Y. B. & Bradley, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminar… more

Date: October 1, 1979

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Fission product release from high gap-inventory LWR fuel under LOCA conditions

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was tes… more

Date: January 1, 1980

Creator: Lorenz, R. A.; Collins, J. L.; Osborne, M. F. & Malinauskas, A. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Developments in sensitivity theory

A review of recent developments in sensitivity theory is presented with an emphasis on (a) extensions to new areas such as thermal hydraulics, reactor depletion, multi-constraint and extrema problems, and (b) recent mathematical refinements to and extensions of the basic theory. The diverse new areas of application are discussed from a unified theoretical viewpoint based on nonlinear functional analysis. Several new applications of sensitivity theory are presented for problems in constrained re… more

Date: January 1, 1980

Creator: Cacuci, D. G.; Greenspan, E.; Marable, J. H. & Williams, M. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Uncertainty in the power distribution for a fast reactor burnup cycle

Demonstration that advanced reactor designs satisfy safety and performance goals requires the analysis of uncertainties in calculated reactor characteristics. Two of the important performance characteristics of advanced liquid metal reactors (LMR) are the burnup reactivity swing and the local power peaking factor. Previous work reported a study of the uncertainty in the burnup reactivity swing attributable to nuclear data uncertainties for a typical advanced LMR design. The results are reported… more

Date: January 1, 1989

Creator: Downar, D.J.; Khalil, H. (Purdue Univ., Lafayette, IN (USA) & Argonne National Lab., IL (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department

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End of life fission product distributions in F-1 experiment fuel rods

Fission product migration and end-of-life distributions were examined in the F-1 (X094) series of sealed, mixed-oxide fuel rods irradiated in the fast flux of EBR-II. Cesium, rubidium, iodine, and strontium data obtained from axial gamma scanning, mass spectrometry, and radiochemical analyses are presented. The results show significant migration of cesium, probably as both a volatile species and as the noble gas precursor. Cesium metal species leaving the fuel region accumulate predominately at… more

Date: May 1, 1980

Creator: Goodin, D. T.; Langer, S. & Bell, W. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Technical Division quarterly progress report, October 1--December 31, 1977. [Fuel cycle research and development; special materials production]

Results are presented on the fluidized-bed calcination of simulated radioactive waste from the reprocessing of spent commercial nuclear fuel and on the removal of actinide elements from the waste prior to calcination. Other programs include the development of storage technology for /sup 85/Kr waste; and the behavior of volatile radionuclides during the combustion of HTGR graphite-based fuel. The long-term management of defense waste from the ICPP covers post-calcination treatment of ICPP calcin… more

Date: February 1, 1978

Creator: Slansky, C.M. (ed.)

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Time series analysis of reactor thermocouple data. [LMFBR]

Time-series analysis techniques are applied to nuclear reactor thermocouple data to investigate coolant temperatures measured within the fueled test assembly. The coolant temperature distribution within a fuel assembly affects the length of time a fuel assembly may be operated in a power reactor and, therefore, is an important economic consideration in the design of reactor fuel systems. Frequency-domain signal conditioning techniques were used to reveal the smoothly varying thermocouple signal… more

Date: April 10, 1980

Creator: Devary, J.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: Irradiation tests. Progress report, July 1--September 30, 1977. [PuO/sub 2/; UO/sub 2/]

Information is presented concerning the examination of fast reactor fuels, FBR analytical quality assurance standards, and analytical methods development irradiation tests.

Date: November 1, 1977

Creator: Baker, R.D. (comp.)

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Sensitivity and uncertainty analysis for functionals of the time-dependent nuclide density field. [ORIGEN-A]

An approach to extend the present ORNL sensitivity program to include functionals of the time-dependent nuclide density field is developed. An adjoint equation for the nuclide field was derived previously by using generalized perturbation theory; the present derivation makes use of a variational principle and results in the same equation. The physical significance of this equation is discussed and compared to that of the time-dependent neutron adjoint equation. Computational requirements for de… more

Date: April 1, 1978

Creator: Williams, M.L. & Weisbin, C.R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Light water reactor fuel reprocessing: dissolution studies of voloxidized fuel

Voloxidation is a proposed head-end process to remove tritium from irradiated LWR fuel by roasting the fuel in the presence of oxygen. The process oxidizes UO/sub 2/ to a fine U/sub 3/O/sub 8/ powder with high surface area for dissolution. Small-scale tests with irradiated Robinson, Oconee, and Saxton reactor fuels have been made to determine the dissolution behavior of both voloxidized and nonvoloxidized (UO/sub 2/) fuel. No significant technical problems were encountered in batch-dissolving o… more

Date: January 1, 1977

Creator: Johnson, D. R. & Stone, J. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Performance potential of (Th,U) carbide and (Th,U) nitride fuel in 1200 MWe LMFBR's

An evaluation of the performance potential of thorium-uranium carbide and nitride fuel in 1200 MW(e) homogeneous and heterogeneous LMFBR's has been completed. Comparisons were done with 9.40 mm outer diameter, 0.38 mm cladding fuel pins at a selected peak (3 sigma) linear power of 98.4 kW/m for all cores. Breeding ratios for the homogeneous (Th,U)C and (Th,U)N cores are 1.12 and 1.16, respectively. In the heterogeneous cores the breeding ratios are 0.08 to 0.09 greater. A comparative breeding r… more

Date: April 1, 1978

Creator: Caspersson, S. A. & Kulwich, M. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fuel systems for compact fast space reactors

About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser po… more

Date: December 1, 1983

Creator: Cox, C.M.; Dutt, D.S. & Karnesky, R.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Irradiated fuel inspection in a storage pond with no fuel movement and an uncollimated detector

It is assumed that the assemblies are not to be moved during the inspection in order to reduce the time required. The task is simplified further when the detector is well collimated. If the collimation is insufficient to respond to only one assembly from among an array of fuel assemblies, then an unfolding process is needed. A formulation of the detector's behavior is given followed by an unfolding process to yield the detector's response to each assembly as if it were isolated. Some simulated … more

Date: July 1, 1980

Creator: Rinard, P. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

The effects of the thermally-induced cracking and subsequent relocation of UO/sub 2/ fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding … more

Date: April 1, 1980

Creator: Williford, R. E.; Mohr, C. L.; Lanning, D. D.; Cunningham, M. E.; Rausch, W. N. & Bradley, E. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fuel performance annual report for 1986

This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs.

Date: March 1, 1988

Creator: Bailey, W.J. & Wu, S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Design of unique pins for irradiation of higher actinides in a fast reactor

The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradi… more

Date: March 1, 1982

Creator: Basmajian, J. A.; Birney, K. R.; Weber, E. T.; Adair, H. L.; Quinby, T. C.; Raman, S. et al.

Item Type: Article

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, 1979

Progress in HTGR studies is reported in the following areas: HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; and evaluation of the pebble-bed HTR.

Date: July 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Light-Water-Reactor Safety Research Program: Quarterly Progress Report, January-March 1980

This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

Date: August 1, 1980

Creator: Massey, Walter E. & Kyger, Jack A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Final Design Analyses of CRBRP Fuel Rod

The key results are presented of the CRBRP fuel rod analyses performed in support of the CRBRP fuel assembly final design review in December, 1977. For the CRBRP fuel rod design environment, it is shown that the life limiting criteria on ductility limited strain (DLS) and cumulative damage function (CDF) are satisfied. The conservatism in these analyses is demonstrated by comparison to the upset enthalpy based failure potential criteria derived from cladding breach experience for TREAT simulate… more

Date: January 1, 1980

Creator: Bishop, B. A. & Vijuk, R. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

End-of-irradiation data report for the instrumented fuel assembly (IFA)-527. [PWR; BWR]

This report presents data obtained during the irradiation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 230-..mu..m diametral gaps and one rod with similar f… more

Date: May 1, 1982

Creator: Cunningham, M. E. & Lanning, D. D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of /sup 235/U. Fuel plates containing 33 v/o U/sub 3/Si and U/sub 3/Si/sub 2/ behaved very well up to this burnup. Plates containing 33 v/o U/sub 3/Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U/sub 3/Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U/sub 3/Si showed an increase swelling rate around 85% burnup. 5 refs., 10 … more

Date: January 1, 1985

Creator: Hofman, G.L. & Neimark, L.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department