scholarly journals Development of Fuel Product Barrier Monitoring System Based on State Functions in State-Oriented Emergency Operating Procedure

2021 ◽  
Vol 2021 ◽  
pp. 1-9
Author(s):  
Wang Zhenying ◽  
Shi Yanming ◽  
Xu Huibo ◽  
Zhang Lijun ◽  
Zhou Shiqing ◽  
...  
Keyword(s):  
Decision Making ◽  
Monitoring System ◽  
Operating Procedure ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Initial Conditions ◽  
Pressurized Water ◽  
Emergency Operating ◽  
State Functions

For pressurized water reactor nuclear power plants, in order to prevent the release of radioactive substances into environment, fission product barriers (FPBs) are constructed based on the concept of defense-in-depth, including fuel clad, reactor coolant system (RCS), and containment; the status of these FPBs is then acting as an important dimension to decision-making of emergency action levels (EALs). For CPR1000 nuclear power plants, state functions defined in state-oriented emergency operating procedure (SOP) are used to characterize postaccident physical conditions; their degradation substantially represents the challenges on fundamental safety functions and then on the integrity of FPBs in like manner, so degradation of these state functions is referred to as determining initial conditions of each FPB, by which the link between SOP and EALs is established. Then, an intelligent FPB monitoring system (FPBMS) aiming to automatically monitor states of FPBs is developed, verified, and validated. The pioneering work, by building bridges between state functions and initial conditions of FPBs and then computerizing them innovatively, proves that dynamical monitoring of states of FPBs during accident evolvement and real-time indication of loss or potential loss of FPBs can be achieved, which is most helpful in decision-making of EALs.

Operating procedure automation to enhance safety of nuclear power plants

1989 ◽  
Vol 110 (3) ◽  
pp. 277-297 ◽  
Author(s):  
Abdo A. Husseiny ◽  
Zeinab A. Sabri ◽  
Dan Packer ◽  
J.W. Holmes ◽  
S. Keith Adams ◽  
...  
Keyword(s):  
Operating Procedure ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

scholarly journals Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity

2018 ◽  
Vol 4 (2) ◽  
pp. 119-125
Author(s):  
Vadim Naumov ◽  
Sergey Gusak ◽  
Andrey Naumov
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Mass Composition ◽  
Published Data ◽  
Pressurized Water ◽  
Fuel Cycles ◽  
Reactor Cores

The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235U, 238U) and accumulation of long-lived fission products (85Kr, 90Sr, 137Cs, 151Sm) and actinoids (238,239,240,241,242Pu, 236U, 237Np, 241Am, 244Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors.

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