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SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 different elements plus a stainless steel

Description: The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables (auth)
Date: January 1, 1976
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

SB3. Experiment on secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for each of 13 different elements plus a stainless steel

Description: The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for iron, oxygen, sodium, aluminum, copper, titanium, calcium, potassium, silicon, nickel, zinc, barium, sulfur, and a type 321 stainless steel. The gamma-ray production cross sections were binned into 0.5-MeV wide gamma-ray energy intervals. 29 tables, 1 figure (auth)
Date: January 1, 1976
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Parameter and fluence-rate covariances in lepricon

Description: The LEPRICON code system is now available from the Radiation Shielding Information Center at ORNL as PSR-277. The system consists of modules that involve both the calculation of neutron fluence rates through PWR pressure vessels and the adjustment of these fluence rates with reduced uncertainties based on surveillance dosimetry. This paper describes in detail the manner in which important parameter uncertainties are partitioned and quantified as part of the input to the adjustment procedure, and to what degree they are applicable to all reactor designs. 7 refs., 1 fig., 1 tab.
Date: January 1, 1990
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Overview of CSEWG shielding benchmark problems

Description: The fundamental philosophy behind the choosing of CSEWG shielding benchmarks is that the accuracy of a certain range of cross section data be adequately tested. The benchmarks, therefore, consist of measurements and calculations of these measurements. Calculations for which there are no measurements provide little information on the adequacy of the data, although they can perhaps indicate the sensitivity of results to variations in data.
Date: January 1, 1979
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

S/sub n/ transport calculations of the PCA experiments with some estimated uncertainties

Description: Detailed flux calculations have been carried out for both the 8/7 and 12/13 configurations involving the Pool Critical Assembly (PCA) at Oak Ridge. In addition, estimates of uncertainties in the calculations arising from both methods approximations and input data uncertainties for two important detector locations in the 8/7 configuration have been made.
Date: January 1, 1980
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Application of the LEPRICON methodology to LWR pressure vessel surveillance dosimetry

Description: A second example of applying the LEPRICON methodology to an existing pressurized water reactor is described. The present application is an analysis of ad hoc dosimetry inserted into the H.B. Robinson-2 reactor to monitor the effects on pressure vessel fluence produced by the introduction of a low-leakage fuel management scheme during cycle 9. The use of simultaneous dosimetry at both a downcomer location and in the reactor cavity allows a quantitative evaluation to be made by the LEPRICON procedure of the relative merits of each location, and the cavity location is found to be superior.
Date: January 1, 1987
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

The LEPRICON (Least-Squares Electric Power Research Institute Consolidation) code system: Consolidation of transport analytical and unfolding procedures in LWR (Light Water Reactor) pressure vessel dosimetry

Description: The LEPRICON (for Least-Square Electric Power Research Institute Consolidation program) code system has been developed over the past ten years to provide a complete analysis of Light Water Reactor (LWR) pressure vessel dosimetry. The system incorporated nine modules. All but one of the modules treat various aspects of neutron transport from the core through the reactor internals to dosimetry locations in the downcomer and/or reactor cavity regions and to critical fluence locations in the pressure vessel. The LEPRICON adjustment module, on the other hand, performs a state-of-the-art least-squares analysis of the results from the transport modules, a procedure often referred as spectral unfolding or the combining of integral and differential data. In terms of development, the adjustment module alone required about 70 percent of the total effort, for reasons that soon will become apparent. The results from the LEPRICON system thus represent prior and adjusted fluences in each of 38 groups from 0.1 to 20 MeV, along with their corresponding standard deviations, at critical locations on the pressure vessel. 7 refs., 1 fig.
Date: January 1, 1988
Creator: Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Comparison of calculations with neutron dosimetry measurements performed at the Oak Ridge Poolside Facility

Description: The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment.
Date: January 1, 1981
Creator: Maerker, R.E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Calculations of the Westinghouse perturbation experiment at the Poolside Facility

Description: Discrete ordinate calculations are made and the results compared with measurements performed in the Poolside Facility for the purpose of validating various procedures adopted for the analysis of this facility. In addition, these calculations can be specifically used to verify the interpretation of measurements made to infer the perturbation effect of a Westinghouse surveillance capsule in a typical radiation environment. Comparisons indicate agreement on an absolute scale between measured and calculated reaction rates to within about 10% and agreement of the perturbation effect to within about 2%.
Date: January 1, 1982
Creator: Maerker, R.E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Calculations of the startup experiments at the Poolside Facility

Description: Discrete ordinate calculations are made and the results compared with measurements performed in the startup experiment at the Poolside Facility. Because of the physical size of the simulated surveillance capsule used in this experiment, the analytic procedure is more complicated than one adopted in earlier calculations of the PCA-PVF and PSF. The comparisons indicate the pressure vessel fluences in the long-term irradiation experiments still presently going on at the PSF, and which are geometrically identical to the startup experiment, can only be predicted to within about 20%.
Date: January 1, 1982
Creator: Williams, M.L. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Accounting for time dependent source variations in surveillance dosimetry analysis

Description: One of the difficulties encountered in the calculation of dosimetry reaction rates is how to account for the time dependent behavior of the core source during the irradiation period. Indeed, even the obtaining of this source time dependence in adequate detail is not a trivial task. The straightforward approach of performing a DOT4 or similar transport calculation for each new relative source distribution, although correct, might be prohibitively expensive and time consuming when the irradiation period spans one or more complete fuel cycles, as it normally does. An alternative approach exists in the generation of a set of adjoint fluxes using DOT4 in the adjoint mode. Equations necessary for using adjoint approach on the Arkansas-1 reactor are presented.
Date: January 1, 1983
Creator: Maerker, R.E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Uncertainties and biases arising from methods approximations: the calculation of reaction rates in the PCA 8/7 configuration

Description: In the determination of the energy spectrum of the neutron fluxes at the surveillance and 1/4 T positions of LWRs by analyzing measured reaction rates, a calculation of these fluxes is both useful and informative in guiding the unfolding, regardless of the particular unfolding procedure used. It should thus be an important part of any unfolding procedure to be able to calculate not only the fluxes but also to estimate both the corrections to this calculation arising from various methods approximations and the uncertainties in these corrections. In at least one unfolding procedure, that utilizing a generalized least squares technique, knowledge of all uncertainties including those arising from calculational methods approximations is an essential part of the input. The particular problem addressed in this paper is the estimation of some correction factors (i.e., biases or bias factors) and their uncertainties arising from various approximations used in calculating the fluxes and reaction rates in the Pool Critical Assembly (PCA) at Oak Ridge for the 8/7 configuration. In addition, the uncertainties in the calculated fluxes and reaction rates arising as the result of uncertainties in the non-nuclear data input was investigated.
Date: January 1, 1979
Creator: Maerker, R.E. & Wagschal, J.J.
Partner: UNT Libraries Government Documents Department

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

Description: A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.
Date: January 1, 1982
Creator: Maudlin, P.J. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Correlations between calculated surveillance dosimeter activities and pressure vessel fluxes in the Arkansas Nuclear One-Unit 1 Reactor

Description: An illustration of the magnitudes of the cross-correlations between cavity dosimetry and fluxes at an important location in the pressure vessel is given for the case of the ANO-1 reactor and provides a correspondence with previously reported flux covariance reduction factors. These correlations are seen to extend over energy regions far wider than would be intuitively expected because of the highly correlated nature of the calculated pressure vessel fluxes themselves, thus enhancing the range of pressure vessel flux information that is provided by the dosimetry. Finally, the high degree of correlation between the cavity dosimeters and pressure vessel fluxes shows that ex-vessel dosimetry can be successfully applied to reactor pressure vessel damage surveillance programs.
Date: January 1, 1985
Creator: Maerker, R.E. & Broadhead, B.L.
Partner: UNT Libraries Government Documents Department

Calculated spectral fluences and dosimeter activities for the metallurgical blind test irradiations at the ORR-PSF. [PWR]

Description: Fluence rate, fluence, and activity calculations were performed for each of the three exposures (two surveillance capsules and a pressure vessel capsule) performed during the two-year metallurgical blind test experiment at the ORR-Poolside Facility in Oak Ridge. Motivation for these calculations was prompted by differences of up to 25% between dosimetry measurements performed in the earlier startup scoping experiment and the two-year experiment. Comparisons of the dosimeter end-of-irradiation activities with HEDL measurements indicate agreement generally within 15% for the first surveillance capsule, 5% for the second, and 10% for three locations in the pressure vessel capsule, which are as good as (if not somewhat better than) comparisons in the startup and the two-year experiments and confirm the presence of a significant cycle-to-cycle variation in the core leakage.
Date: January 1, 1984
Creator: Maerker, R.E. & Worley, B.A.
Partner: UNT Libraries Government Documents Department

Investigation of steel--sodium--iron shields

Description: An analysis of experimental data from 21 fast reactor shield configurations containing steel, sodium, and iron were made as part of a study of the upper axial shielding needs of the Clinch River Breeder Reactor. The measured data were analyzed using both one- and two-dimensional discrete ordinates transport codes and several cross section libraries based on ENDF/B-IV data with group structures of 51 and 171 neutron groups. One-dimensional sensitivity studies using the 171 group library and ENDF/B-IV covariance files for sodium and iron data were used to determine the sensitivities of the measured data to multigroup cross sections and to estimate uncertainties in the calculated results. Results indicate that the standard 51-group design cross section library could be expected to predict the measurements to within 30% over 12 decades of attenuation although a few of the deepest penetration configurations showed disagreements as large as a factor of three. The sensitivity results revealed very high sensitivity of the measurements to total cross section minima and cross sections from 5 to 10 MeV in sodium and iron in the deep penetration configurations. As a result, large uncertainties in the calculated results arose from small uncertainties in the cross section data. These results indicate the need for better measurements of the total cross section minima in sodium, especially around 300 keV.
Date: January 1, 1978
Creator: Oblow, E.M. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department

Sensitivities of the flux spectrum in the cavity of a PWR to variations in the core source distribution

Description: As a part of an ongoing, EPRI-sponsored project whose aim is the quantification and reduction of fluence uncertainties in the pressure vessel of operating PWR's, this work describes the calculation of sensitivities necessary for the propagation of PWR core source distribution uncertainties to the flux spectrum at locations of interest (e.g., the cavity or T/4 pressure vessel locations) in the AN0l reactor. In this case standard perturbation theory requires an adjoint run to be made for each group flux since each group flux is a response. An alternate approach has been developed by Cacuci which should be more efficient than the standard approach although it has not yet been applied to a flux spectrum response.
Date: June 3, 1984
Creator: Broadhead, B.L. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department