RAMA Surveillance Capsule and Component Activation Analyses
- TransWare Enterprises Inc., 1565 Mediterranean Dr., Sycamore, IL 60178 (United States)
- Electric Power Research Institute, 1300 West W. T. Harris Blvd., Charlotte, NC 28262 (United States)
This paper presents the calculated-to-measured ratios associated with the application of the RAMA Fluence Methodology software to light water reactor surveillance capsule and reactor component activation evaluations. Comparisons to measurements are performed for pressurized water reactor and boiling water reactor surveillance capsule activity specimens from seventeen operating light water reactors. Comparisons to measurements are also performed for samples removed from the core shroud, top guide, and jet pump brace pads from two reactors. In conclusion: The flexible geometry modeling capabilities provided by RAMA, combined with the detailed representation of operating reactor history and anisotropic scattering detail, produces accurate predictions of the fast neutron fluence and neutron activation for BWR and PWR surveillance capsule geometries. This allows best estimate RPV fluence to be determined without the need for multiplicative bias corrections. The three-dimensional modeling capability in RAMA provides an accurate estimate of the fast neutron fluence for regions far removed from the core mid-plane elevation. The comparisons to activation measurements for various core components indicate that the RAMA predictions are reasonable, and notably conservative (i.e., C/M ratios are consistently greater than unity). It should be noted that in the current evaluations, the top and bottom fuel regions are represented by six inch height nodes. As a result, the leakage-induced decrease in power near the upper and lower edges of the core are not well represented in the current models. More precise predictions of fluence for components that lie above and below the core boundaries could be obtained if the upper and lower fuel nodes were subdivided into multiple axial regions with assigned powers that reflect the neutron leakage at the top and bottom of the core. This use of additional axial sub-meshing at the top and bottom of the core is analogous to the use of pin-wise meshing in peripheral bundles to accurately represent radial leakage effects. The representation of thermal neutron fluence and activations are found to be reasonably accurate and consistently conservative, as demonstrated by comparison to the reactor component thermal neutron reaction activation measurements. Further improvement in the comparisons to measurements could be achieved by exploring the impact of enhanced sub-meshing of the geometry near the components of interest. The mesh densities utilized in the current evaluation are consistent with the mesh requirements for high energy neutron transport. The substantially shorter transport lengths for thermal neutrons relative to high energy neutrons suggests that localized regions of finer meshing are needed in the vicinity of those reactor components requiring thermal neutron fluence evaluations. (authors)
- Research Organization:
- American Society for Testing and Materials - ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA, 19428-2959 (United States); European Working Group on Reactor Dosimetry - EWGRD, SCK.CEN, Mol (Belgium)
- OSTI ID:
- 22086965
- Report Number(s):
- INIS-US-13-ISRD-14-os6p-5; TRN: US13V0006045412
- Resource Relation:
- Conference: ISRD-14: 14. International Symposium on Reactor Dosimetry, Bretton Woods, NH (United States), 22-27 May 2011; Other Information: Country of input: France; 2 refs.
- Country of Publication:
- United States
- Language:
- English
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