Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008

This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2008 annual reports submitted by five of the seven categories1 of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Because there are no geologic repositories for high-level waste currently licensed and no low-level waste disposal facilities in operation, only five categories will be considered in this report

High level seismic/vibrational tests at the HDR: An overview

As part of the Phase II testing at the HDR Test Facility in Kahl/Main, FRG, two series of high-level seismic/vibrational experiments were performed. In the first of these (SHAG) a coast-down shaker, mounted on the reactor operating floor and capable of generating 1000 tonnes of force, was used to investigate full-scale structural response, soil-structure interaction (SSI), and piping/equipment response at load levels equivalent to those of a design basis earthquake. The HDR soil/structure system was tested to incipient failure exhibiting highly nonlinear response. In the load transmission from structure to piping/equipment significant response amplifications and shifts to higher frequencies occurred. The performance of various pipe support configurations was evaluated. This latter effort was continued in the second series of tests (SHAM), in which an in-plant piping system was investigated at simulated seismic loads (generated by two servo-hydraulic actuators each capable of generating 40 tonnes of force), that exceeded design levels manifold and resulted in considerable pipe plastification and failure of some supports (snubbers). The evaluation of six different support configurations demonstrated that proper system design (for a given spectrum) rather than number of supports or system stiffness is essential to limiting pipe stresses. Pipe strains at loads exceeding the design level eightfold were still tolerable, indicating that pipe failure even under extreme seismic loads is unlikely inspite of multiple support failures. Conservatively, an excess capacity (margin) of at least four was estimated for the piping system, and the pipe damping was found to be 4%. Comparisons of linear and nonlinear computational results with measurements showed that analytical predictions have wide scatter and do not necessarily yield conservative responses, underpredicting, in particular, peak support forces

Occupational radiation Exposure at Agreement State-Licensed Materials Facilities, 1997-2010

The purpose of this report is to examine occupational radiation exposures received under Agreement State licensees. As such, this report reflects the occupational radiation exposure data contained in the Radiation Exposure Information and Reporting System (REIRS) database, for 1997 through 2010, from Agreement State-licensed materials facilities

Reactor safety research programs : quarterly reports /

v. 1- January-March 1983, v.2- April-June 1983, v.3- July-September 1983, v.4- October-December 1983.Performing organization : Pacific Northwest Laboratory.Mode of access: Internet

Subsystem response review : seismic safety margins research program /

"UCRL-15215.""NUREG/CR-1700.""NRC FIN no. A0126.""Date published: January 1981.""Lawrence Livermore Laboratory.""Nuclear Services Corporation."Includes bibliographical references.Mode of access: Internet

Analysis of the Core Exit Temperature and the Peak Cladding Temperature during a SBLOCA. Application to a scaled-up model

[EN] During Loss-Of-Coolant Accidents (LOCA), operators may start Accident Management (AM) actions when the Core Exit Temperature (CET) measured by thermocouples exceeds a certain value. However, a significant time delay and temperature discrepancy in the superheat detection was observed in several facilities. This work is focused on clarifying CET thermocouple responses versus Peak Cladding Temperature (PCT) and to study if the same physical phenomena are reproduced in two TRACE5 models with different geometry (a Large Scale Test Facility (LSTF) and a scaled-up LSTF) during a pressure vessel upper head Small Break LOCA (SBLOCA). Results obtained show that the delay between the core uncover and the CET excursion is reproduced in both cases.The authors are grateful to the Management Board of the OECD-NEA ROSA Project; thus, this work contains findings produced within this project. This work is partially supported by the Grant-in-Aid for Scientific Research of the Spanish Ministerio de Educacion (Grant number: AP2009-2600), the Spanish Ministerio de Ciencia e Innovacion under Projects ENE2011-22823 and ENE2012-34585, and the Generalitat Valenciana under Projects PROMETEOII/2014/008 and ACOMP/2013/237.Querol, A.; Gallardo Bermell, S.; Verdú Martín, GJ. (2016). Analysis of the Core Exit Temperature and the Peak Cladding Temperature during a SBLOCA. Application to a scaled-up model. Journal of Nuclear Engineering and Radiation Science. 2(2):1-6. https://doi.org/10.1115/1.4031016S1622Tóth, I., Prior, R., Sandervag, O., Umminger, K., Nakamura, H., Muellner, N., Cherubini, M., Del Nevo, A., D’Auria, F., Dreier, J., Alonso, J. R., and Amri, A., 2010, “Core Exit Temperature (CET) Effectiveness in Accident Management of Nuclear Power Reactors,” Committee on the Safety of Nuclear Installations, OECD, Nuclear Energy Agency, Tech. Rep. NEA/CSNI/R(2010)9.Freixa, J., Martínez-Quiroga, V., Zerkak, O., & Reventós, F. (2015). Modelling guidelines for core exit temperature simulations with system codes. Nuclear Engineering and Design, 286, 116-129. doi:10.1016/j.nucengdes.2015.02.003Adams, J. P., and McCreery, G. E., 1983, “Detection of Inadequate Core Cooling With Core Exit Thermocouples: LOFT PWR Experience,” U.S. Nuclear Regulatory Commission, Washington, NUREG/CR-3386.Suzuki, M., 1993, “Characteristic Responses of Core Exit Thermocouples During Inadequate Core Cooling in Small Break LOCA Experiments Conducted at LSTF of ROSA-IV Program,” Proceedings of ICONE2, San Francisco, CA, American Society of Mechanical Engineers, United Engineering Center, New York, Vol. 1, pp. 63–68.SUZUKI, M., & NAKAMURA, H. (2010). Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF. Journal of Nuclear Science and Technology, 47(12), 1193-1205. doi:10.1080/18811248.2010.9720986Thermalhydraulic Safety Research Group, Nuclear Safety Research Center, 2006, “Final Data Report of OECD/NEA ROSA Project Test 6-1 (1.9% Pressure Vessel Upper-Head Small Break LOCA Experiment SB-PV-09 in JAEA),” Japan Atomic Energy Agency, Tokai-mura, Japan.The ROSA-V Group, 2003, “ROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies,” JAERI-Tech, Tokai-mura.Gallardo, S., Abella, V., and Verdú, G., 2010, “Assessment of TRACE 5.0 Against ROSA Test 6-1, Vessel Upper Head SBLOCA,” U.S. Nuclear Regulatory Commission, Washington, NUREG/IA-0245.Freixa, J., & Manera, A. (2010). Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE. Nuclear Engineering and Design, 240(7), 1779-1788. doi:10.1016/j.nucengdes.2010.02.007Freixa, J., & Manera, A. (2011). Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test. Nuclear Engineering and Design, 241(3), 888-896. doi:10.1016/j.nucengdes.2010.12.016Queral, C., González-Cadelo, J., Jimenez, G., & Villalba, E. (2011). Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed. Nuclear Technology, 175(3), 572-593. doi:10.13182/nt11-a12507Queral, C., González-Cadelo, J., Jimenez, G., Villalba, E., and Perez, J., 2013, “Simulation of LSTF Upper Head Break (OECD/NEA ROSA Test 6.1) With TRACE Code. Application to a PWR NPP Model,” U.S. Nuclear Regulatory Commission, Washington, NUREG/IA-0426.Querol, A., Gallardo, S., & Verdú, G. (2015). Simulation of a SBLOCA in a hot leg. Scaling considerations and application to a nuclear power plant. Nuclear Engineering and Design, 283, 81-99. doi:10.1016/j.nucengdes.2014.10.006Liu, T.-J., Lee, C.-H., & Way, Y.-S. (1997). IIST and LSTF counterpart test on PWR station blackout transient. Nuclear Engineering and Design, 167(3), 357-373. doi:10.1016/s0029-5493(96)01302-7D’Auria, F., & Galassi, G. M. (2010). Scaling in nuclear reactor system thermal-hydraulics. Nuclear Engineering and Design, 240(10), 3267-3293. doi:10.1016/j.nucengdes.2010.06.010Petelin, S., Mavko, B., Končar, B., & Hassan, Y. A. (2007). Scaling of the Small-Scale Thermal-Hydraulic Transient to the Real Nuclear Power Plant. Nuclear Technology, 158(1), 56-68. doi:10.13182/nt07-a382

Light-water-reactor safety materials engineering research programs : quarterly progress report. /

"Prepared for the Division of Engineering Technology, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission under interagency agreement DOE 40-550-75.""Distribution code: R5.""ANL-84-60.""NUREG/CR-3998."Volume 1. January-March 1984 -- volume 2. April-June 1984 -- volume 3. October-December 1984.Mode of access: Internet

Light-water-reactor safety research program.

Format not distributed to depository libraries."NUREG/CR."Mode of access: Internet.Vols. for prepared for the Division of Reactor Safety Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission.Description based on: Oct.-Dec. 1977

Load combination program progress report no. 6. /

"UCID-18674, Vol. 2.""NUREG/CR-1624, Vol. 2.""NRC FIN no. A0133, A0362, A0363"--Vol. 2."Date published: May 1981"--Vol. 2."Lawrence Livermore Laboratory"--Vol. 2.Mode of access: Internet.Description based on v. 2