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Uncertainties in the estimation of radiation-damage parameters

Description: The evaluation of radiation embrittlement and the determination of safety limits requires the knowledge of uncertainties in the estimation of radiation exposure parameters like flux greater than 1.0 MeV or dpa of iron. Least squares adjustment methods can be used for the estimation of the exposure parameters and their undertainties. It is of interest to determine how the uncertainties are influenced by the input data, in particular, the selection of dosimeters for the determination of exposure parameters. This investigation is simplified by the fact that in least squares methods the output uncertainties depend primarily on the input uncertainties and very little on the measured values themselves. Thus, the expected uncertainties can be determined without actually making measurements. In this paper, uncertainties for exposure parameters are calculated for a variety of foil sets. The consequences of this investigation for surveillance dosimetry are discussed.
Date: January 1, 1982
Creator: Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

Design and use of the Embrittlement Data Base (EDB)

Description: The architecture of the Embrittlement Data Base (EDB) is described. This data base contains a comprehensive collection of experimental data related to irradiations of reactor pressure vessel steels in surveillance capsules and test reactors. Software is being developed for easy retrieval and analysis of the data. Data and software will be made available to interested parties on a cooperative basis.
Date: January 1, 1987
Creator: Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

LSL-M1 and LSL-M2: two extensions of the LSL adjustment procedure for including multiple spectrum locations

Description: Most current adjustment procedures, including LSL, can adjust only one spectrum with dosimetry located at the point of the input spectrum. Many radiation experiments have dosimetry at more than one location, and fluence or damage exposure values are desired for locations other than those covered by dosimetry. Thus, the use of single-spectrum dosimetry to these experiments causes considerable loss of information and introduces large uncertainties. Two extensions of the LSL code to cover multiple-spectra adjustment are discussed. Each extension has different restrictions and covers a different range of applications.
Date: January 1, 1984
Creator: Stallmann, F. W.
Partner: UNT Libraries Government Documents Department

ORNL evaluation of the ORR-PSF metallurgical experiment and blind test

Description: A methodology is described to evaluate the dosimetry and metallurgical data from the two-year ORR-PSF metallurgical irradiation experiment. The first step is to obtain a three-dimensional map of damage exposure parameter values based on neutron transport calculations and dosimetry measurements which are obtained by means of the LSL-M2 adjustment procedure. Metallurgical test data are then combined with damage parameter, temperature, and chemistry information to determine the correlation between radiation and steel embrittlement in reactor pressure vessels including estimates for the uncertainties. Statistical procedures for the evaluation of Charpy data, developed earlier, are used for this investigation. The data obtained in this investigation provide a benchmark against which the predictions of the PSF Blind Test can be compared. The results of this investigation and the Blind Test comparison are discussed.
Date: January 1, 1984
Creator: Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

Neutron spectral characterization of the NRC-HSST experiments

Description: Irradiation experiments are being performed for the US Nuclear Regulatory Commission (NRC) Heavy Section Steel Technology (HSST) program. Results of dosimetry performed in the second experiment have been previously reported. Similar procedures were followed in the third experiment. The experiences gained in these two experiments have led to modifications in the composition and distribution of foil dosimeters which monitor the neutron flux-spectra in the irradiated steel specimens. It is expected that in the new experiments much higher accuracies than previously possible can be achieved in the determination of irradiation damage parameters.
Date: January 1, 1979
Creator: Stallmann, F.W. & Kam, F.B.K.
Partner: UNT Libraries Government Documents Department

Foil activation dosimetry at energies below 1 MeV. [Theoretical analysis]

Description: The foil activation method is usually applied to obtain the neutron spectrum at energies above 1 MeV. However, potentially useful information can be gained from detectors whose main responses lie in the low-energy region. Efficient quantitative methods are developed to extract this information using ''window function'' and linear programming techniques. This method is restricted essentially to energies below 1 keV. In the 1 keV to 1 MeV regions, which is important for damage analysis, only a combination of foil detector methods and neutron transport calculations can provide satisfactory results.
Date: January 1, 1977
Creator: Stallmann, F.W. & Kam, F.B.K.
Partner: UNT Libraries Government Documents Department

Reactor calculation benchmark PCA blind test results

Description: Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.
Date: January 1, 1980
Creator: Kam, F.B.K. & Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

LWR surveillance dosimetry improvement program: PSF metallurgical blind test results

Description: The metallurgical irradiation experiment at the Oak Ridge Research Reactor Poolside Facility (ORR-PSF) was designed as a benchmark to test the accuracy of radiation embrittlement predictions in the pressure vessel wall of light water reactors on the basis of results from surveillance capsules. The PSF metallurgical Blind Test is concerned with the simulated surveillance capsule (SSC) and the simulated pressure vessel capsule (SPVC). The data from the ORR-PSF benchmark experiment are the basis for comparison with the predictions made by participants of the metallurgical ''Blind Test''. The Blind Test required the participants to predict the embrittlement of the irradiated specimen based only on dosimetry and metallurgical data from the SSC1 capsule. This exercise included both the prediction of damage fluence and the prediction of embrittlement based on the predicted fluence. A variety of prediction methodologies was used by the participants. No glaring biases or other deficiencies were found, but neither were any of the methods clearly superior to the others. Closer analysis shows a rather complex and poorly understood relation between fluence and material damage. Many prediction formulas can give an adequate approximation, but further improvement of the prediction methodology is unlikely at this time given the many unknown factors. Instead, attention should be focused on determining realistic uncertainties for the predicted material changes. The Blind Test comparisons provide some clues for the size of these uncertainties. In particular, higher uncertainties must be assigned to materials whose chemical composition lies outside the data set for which the prediction formula was obtained. 16 references, 14 figures, 5 tables.
Date: January 1, 1984
Creator: Kam, F.B.K.; Maerker, R.E. & Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

WINDOWS: a program for the analysis of spectral data foil activation measurements

Description: The computer program WINDOWS together with its subroutines is described for the analysis of neutron spectral data foil activation measurements. In particular, the unfolding of the neutron differential spectrum, estimated windows and detector contributions, upper and lower bounds for an integral response, and group fluxes obtained from neutron transport calculations. 116 references. (JFP)
Date: December 1, 1978
Creator: Stallmann, F.W.; Eastham, J.F. & Kam, F.B.K.
Partner: UNT Libraries Government Documents Department

Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

Description: An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed.
Date: January 1, 1983
Creator: Williams, M.L.; Stallmann, F.W.; Maerker, R.E. & Kam, F.B.K.
Partner: UNT Libraries Government Documents Department

NRC data base for power reactor surveillance programs and for irradiation experiments results

Description: The radiation damage of pressure vessel materials in nuclear reactors depends on many different factors, primarily fluence, fluence spectrum, fluence rate, irradiation temperature, and chemistry. These factors and, possibly, others such as heat treatment and type of flux used in weldments must be considered to reliably predict the pressure vessel embrittlement and to assure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters, low-leakage fuel management, possible life extension, and the need for annealing of the pressure vessel. Large numbers of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. The US Nuclear Regulatory Agency has, therefore, sponsored a project to construct an Embrittlement Data Base (EDB) for a comprehensive collection of data concerning changes in material properties of pressure vessel steels due to neutron irradiation. A first version containing data from surveillance capsules of commercial power reactors, the Power Reactor Embrittlement Data Base (PR-EDB) Version 1, has been completed and is available to authorized users from the Radiation Shielding Information Center at the Oak Ridge National Laboratory. This document provides a discussion of the features of the current database. 1 fig.
Date: January 1, 1991
Creator: Kam, F.B.K. & Stallmann, F.W.
Partner: UNT Libraries Government Documents Department

Power reactor embrittlement data base

Description: Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs.
Date: January 1, 1989
Creator: Kam, F.B.K.; Stallmann, F.W. & Wang, J.A.
Partner: UNT Libraries Government Documents Department

Neutron-exposure parameters for the fourth HSST series of metallurgical irradiation capsules

Description: The neutron exposure parameters for the Heavy Section Steel Technology (HSST) Experiments performed at the Oak Ridge National Laboratory (ORNL) can be determined conservatively to +-10% (1sigma) variance. The neutron exposure parameters used for this study were fluence greater than 1 MeV, fluence greater than 0.1 MeV, and displacements per atom (dpa). Measured reaction rates, calculated neutron transport fluxes, and cross sections values were combined in the logarithmic least square adjustment code LSL.
Date: January 1, 1982
Creator: Kam, F.B.K.; Stallmann, F.W.; Baldwin, C.A. & Fabry, A.
Partner: UNT Libraries Government Documents Department

PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

Description: Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.
Date: January 1, 1994
Creator: Stallmann, F. W.; Wang, J. A.; Kam, F. B. K. & Taylor, B. J.
Partner: UNT Libraries Government Documents Department

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Description: The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.
Date: January 1, 1994
Creator: Stallmann, F. W.; Wang, J. A. & Kam, F. B. K.
Partner: UNT Libraries Government Documents Department