Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant
Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.
Design and development of a comprehensive program for the assessment and analysis of environmental effects due to the release of radioactive materials from the stack of nuclear installations: A case study in Bushehr nuclear power plant.
The current investigation presents a comprehensive program called KIANA to assess and analyze the environmental effects due to the release of radioactive materials from the stack of nuclear installations. Bushehr Nuclear Power Plant Unit One (BNPP-1) is modeled using the KIANA software and the release of radioactive materials from the BNPP-1 stack on the members of the public through the airborne pathway is evaluated during normal operation as well as in the accident conditions including Design Basis Accidents (DBA) and Beyond Design Basis Accidents (BDBA). To verify and validate the KIANA software, the results of the KIANA are compared with the results of DOZA_M, ESTE, PC-CREAM 98, RECASS Express, and the values of the total effective dose monitored by the local detectors of BNPP-1. The KIANA is developed based on the Gaussian diffusion model written using C# programming language. In the current research, the total effective dose received by a member of the public due to the radioactive plume passage through the airborne pathway is calculated in the normal condition of BNPP-1. Moreover, the total effective dose in the case of the primary-to-secondary leakage inside the steam generators, the total effective dose in the case of Large Break Loss of Coolant Accident (LBLOCA) and Small Break Loss of Coolant Accident (SBLOCA), and the equivalent dose of the thyroid gland tissue for an infant group (1-8 years old) in the case of LBLOCA in DBA conditions are evaluated. Finally, the absorbed dose of the whole body of adults at the initial stage after a BDBA, absorbed dose of the thyroid gland tissue for an infant group (1-8 years old) at the initial stage after a BDBA, and the total effective dose in the first year after the accident in the case of a BDBA are assessed. The results of the KIANA software indicate a good agreement with the results of the DOZA_M, ESTE, PC-CREAM 98, RECASS Express computer programs, and the values of the total effective dose monitored by the local detectors of BNPP-1. The developed software has the potential of calculations of concentration and radionuclide dose received through all the exposure pathways such as airborne, foodstuff, marine, soil, animals, and vegetation without any restriction in normal and accident conditions, simultaneously.
The Methodology for Assessing the Radiological Consequences of Routine Releases of Radionuclides to the Environment Used in PC-CREAM 08
This report describes the methodology that has been implemented in the software PCCREAM 08 to assess the radiological impact of routine discharges of radioactive material into the environment. PC-CREAM 08 has been developed by the Health Protection Agency (HPA), with permission from the European Commission (EC), and is an updated version of the EC code PC-CREAM 98. This methodology is closely based on that developed by a number of EU organisations for the EC and which was published in 1995 (RP 72); this was in turn a further development of an original methodology developed for the EC and published in 1979. Although primarily developed for application in Western Europe, a generalised approach has been adopted so that some of the models and methods are appropriate for wider use. Default values are given for many parameters and these have been used to determine illustrative results. The models adopted in the methodology are those considered appropriate for routine releases, ie, releases that can be considered as continuous and constant. Radiological impact assessments involve the calculation of radiation exposures to both individuals and population groups. In the absence of measurements this can is achieved through modelling. The models described in this report predict the transfer of radionuclides in the environment, the pathways by which people may be exposed to radiation and the resulting radiation doses received. The radiological consequences of routine releases of radionuclides are determined using the framework of the system of radiological protection recommended by the International Commission on Radiological Protection (ICRP). The most recent recommendations of ICRP, issued in publication 103, have been taken into account in developing this methodology. However, it should be noted that dose coefficients based on revised radiation and tissue weighting factors have yet to be published and therefore these are taken from publication 60. Acknowledgment and thanks are given to the many individuals and organisations who have contributed to this report and those from which it has been derived.