William N. McElroy

Application of solid state track recorder neutron dosimetry for three mile island unit 2 reactor recovery

Abstract Application of neutron dosimetry for assessment of fuel distribution throughout the Three Mile Island-2 (TMI-2) reactor core region and the primary coolant system is advanced. Neutron dosimetry in the reactor cavity, i.e. the cavity between the pressure vessel and the biological shield, could provide data for the assessment of the core fuel distribution. A more immediate task entails locating and quantifying the amount of fuel debris in the ex-core primary coolant system in the range of 1 to 1000 kg. Solid state track recorder (SSTR) neutron dosimetry is considered for such exploratory scoping experiments at TMI-2. The sensitivity of mica- 2 3 5 U (asymptotically thick) SSTR has been ascertained for such environments. For plausible geometric assumptions and environmental conditions, it has been demonstrated that the SSTR method has adequate sensitivity to properly respond and detect fuel quantities of the order of 1 kg in the ex-core primary coolant system.

Trend Curve Data Development and Testing

Existing trend curves do not account for previous and more recently observed test and power reactor flux-level, thermal neutron and ..gamma..-ray field-induced effects. Any agreement between measured data and trend curve predictions that does not adequately represent the important neutron environmental and temperature effects as well as the microstructural damage processes, therefore, could be fortuitous. Empirically derived end-of-life (EOL) and life-extension-range (LER) trend curves are presented and discussed in this paper for high temperature (approx.288/sup 0/C (550/sup 0/F)) irradiation of two weld, two plate, and two forging pressure vessel (PV) steels and low-temperature (approx.60/sup 0/C (140/sup 0/F)) irradiation of one support structure-type steel.

Light water reactor pressure vessel surveillance using reactor cavity solid state track recorder neutron dosimetry

Abstract Solid State Track Recorder (SSTR) Neutron Dosimeters have been developed for use in power reactors to provide information on the cumulative neutron dose received by the reactor pressure vessel during operation. The accumulation of neutron dose by the pressure vessel results in radiation damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, the results of dosimeter measurements are evaluated and used to estimate the extent of radiation damage. Among the requirements for SSTR neutron dosimetry are high accuracy and ability to provide useful data at high neutron fluences. To this end, ultra low-mass fissionable deposit preparation techniques have been developed, and the absolute accuracies of the measurements have been maintained at the 3–5% level. The status of the deployment of SSTR dosimetry capsules in the reactor cavity region of operating power reactors will be summarized.

Characterization of Fuel Distribution in the Three Mile Island Unit 2 (TMI-2) Reactor System by Neutron and Gamma-Ray Dosimetry

Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. Results obtained from these gamma- ray and neutron dosimetry experiments were 1.3 ± 0.6 kg and 1.7 ± 0.6 kg, respectively, for the fuel content of TMI-2 Demineralizer A.

Neutron dosimetry with solid-state track recorders in the three-mile island unit-2 reactor cavity

Abstract Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit-2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity.

Radiation-induced embrittlement in light water reactor pressure vessels

Abstract In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations. Recent LWR-PV benchmark experiments are analyzed. On this basis, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by this simple exponential behavior are explained and trend-curve models for the prediction of PV embrittlement are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence.

Trend Curve Exposure Parameter Data Development and Testing

An important aspect of the Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) is the effort to develop and test trend curve exposure parameter data. Progress in these trend curve-data correlation analysis activities at HEDL is described. The exposure parameters of primary interest are those associated with the production of displaced atoms and helium in different PV steels, particularly, A302B, A533B, and A508. In order to determine the effect(s) of helium, the production of helium from boron (n,a), iron (n,a), nickel (n,a) and low concentrations of impurity elements are being investigated. Preliminary results of these HEDL investigations and related studies suggest that both displaced iron atoms and total helium content need to be considered in the future development, testing, and application of revisions of selected ASTM standards and Reg. Guide 1.99 and for the generation of PWR and BWR plant specific trend curves.

ASTM Standards Associated with PWR and BWR Power Plant Licensing, Operation and Surveillance

This paper considers ASTM Standards that are available, under revision, and are being considered in support of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Nuclear Power Plant (NPP) licensing, regulation, operation, surveillance and life attainment. The current activities of ASTM Committee E10 and its Subcommittees E10.02 and current activities of ASTM Committee E10 and its Subcommittees E10.02 and E10.05 and their Task Groups (TG) are described. A very important aspect of these efforts is the preparation, revision, and balloting of standards identified in the ASTM E706 Standard on Master Matrix for Light Water Reactor (LWR) Pressure Vessel (PV) Surveillance Standards. The current version (E706-87) of the Master Matrix identifies 21 ASTM LWR physics-dosimetry-metallurgy standards for Reactor Pressure Vessel (RPV) and Support Structure (SS) surveillance programs, whereas, for the next revision 34 standards are identified. The need for national and international coordination of Standards Technology Development, Transfer and Training (STDTT) is considered in this and other Symposium papers that address specific standards related physics-dosimetry-metallurgy issues. 69 refs.

Reaction rate measurements for nuclear reactor analyses and critically data

Abstract Neutron reaction rate measurements with solid state track recorders (SSTR) and radiometric (RM) neutron dosimeters have been conducted at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML) as part of the U.S. Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation of Japan (PNC) Critically Data Development Program. These reaction rate measurements represent benchmark data that can rigorously test the adequacy of neutron transport calculations performed for nuclear reactor analyses as well as for critically safety assessments. In the 220 series of experiments, fast test reactor (FTR) plutonium fuel pins were assembled in a 0.761 cm square lattice array, which was immersed in an organic moderator. In-fuel and in-moderator dosimetry measurements were conducted near axial midplane. To obtain results at a single axial location, corrections were applied for the spatial variation (axial buckling) of the reaction rates. For the in-fuel measurements, dosimeters were placed between fuel pellets thereby creating gaps in the fuel pin column. Consequently, for these in-fuel measurements, the gap-perturbation effect was measured so that reaction rate data could be corrected to zero fuel gap, i.e. the reaction rates in the unperturbed fuel. Experimental uncertainties range from a low of 2–3 percent for U-238 and Th-232 fission rates to a high of 15–16 percent for U-238 capture rates. The uncertainty in the U-238 (n, Σ) moderator results is dominated by the uncertainty in the neutron self-shielding correction factor, which has been estimated to be approximately 15 percent. Uncertainties in U-235, Np-237 and Pu-239 fission rates range from approximately 4 to 7 percent. These latter uncertainties are larger than uncertainties normally achievable in SSTR neutron dosimetry and reflect the fact that the quality of the SSTR electrodeposits prepared for these isotopes was not completely satisfactory. Least-squares spectral analyses of these data were performed with the FERRET-SAND II computer code. These analyses confirm the general consistency of the experimental data and furnish absolute neutron fluxes with assigned uncertainties.

Neutron dosimetry in the three-mile island unit 2 reactor cavity with solid-state track recorders

Abstract Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI - 2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity.