Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

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Preliminary evaluation of the effect of several management measures for the station blackout severe accident of the pressurized water reactor nuclear power plant

Proceedings of International Conference on Information Systems for Crisis Response and Management (ISCRAM) ◽

10.1109/iscram.2011.6184040 ◽

2011 ◽

Author(s):

Wen-kui Lu ◽

Bo Zhao

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Severe Accident ◽

Preliminary Evaluation ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Management Measures ◽

Station Blackout

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Radioactive Source Specification of Bushehr’s VVER-1000 Spent Fuels

Science and Technology of Nuclear Installations ◽

10.1155/2016/4579738 ◽

2016 ◽

Vol 2016 ◽

pp. 1-4 ◽

Cited By ~ 2

Author(s):

Mahdi Rezaeian ◽

Jamshid Kamali

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Initial Step ◽

Total Activity ◽

Radioactive Source ◽

Pressurized Water Reactor ◽

Spent Fuel ◽

Pressurized Water

Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressurized water reactor, was determined. For the depletion and decay calculations, ORIGEN code was utilized. The results are presented for burnups of 30 to 49 GWd/MTHM and different cooling times up to 100 years. According to these results, total activity of a spent fuel assembly with initial enrichment of 3.92%, burnup of 49 GWd/MTHM, and cooling time of 3 years is 1.92 × 1016 Bq. The results can be utilized specifically in transportation/storage cask design for spent fuel management of Bushehr nuclear power plant.

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Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel

10.1115/icone28-63314 ◽

2021 ◽

Author(s):

Jin Feng Huang

Keyword(s):

Thermal Conductivity ◽

Silicon Carbide ◽

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Carbide Matrix ◽

Silicon Carbide Matrix

Abstract After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR. The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR. We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity. The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor. In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.

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The Impact of Climate Changes on the Thermal Performance of a Proposed Pressurized Water Reactor: Nuclear-Power Plant

International Journal of Nuclear Energy ◽

10.1155/2014/793908 ◽

2014 ◽

Vol 2014 ◽

pp. 1-7 ◽

Cited By ~ 4

Author(s):

Said M. A. Ibrahim ◽

Mohamed M. A. Ibrahim ◽

Sami. I. Attia

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Thermal Performance ◽

Nuclear Power ◽

Cooling Water ◽

Plant Performance ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

The Impact

This paper presents a methodology for studying the impact of the cooling water temperature on the thermal performance of a proposed pressurized water reactor nuclear power plant (PWR NPP) through the thermodynamic analysis based on the thermodynamic laws to gain some new aspects into the plant performance. The main findings of this study are that an increase of one degree Celsius in temperature of the coolant extracted from environment is forecasted to decrease by 0.39293 and 0.16% in the power output and the thermal efficiency of the nuclear-power plant considered, respectively.

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A Study About Safety I&C System Software V&V in Nuclear Power Plant

Volume 1: Operations and Maintenance, Aging Management and Plant Upgrades; Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory; Plant Systems, Structures, Components and Materials; I&C, Digital Controls, and Influence of Human Factors ◽

10.1115/icone24-60125 ◽

2016 ◽

Cited By ~ 3

Author(s):

Pengfei Gu ◽

Shengchao Wang ◽

Weihua Chen ◽

Suyuan Yu

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Software Verification ◽

Third Party ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Safety Function ◽

Software Classification

Since the digital instrument and control (I&C) system was used in nuclear power plant (NPP), software Verification and Validation (V&V) is becoming more and more important for ensuring the safety function to be correctly implemented by the safety system. According to the classification of different safety functions of I&C system, with the use of digital technology, software classification needs further discussion to determine. And different classification of software performs different testing. Software V&V processes can be divided into different stages by reference to the standard IEEE 1012-2004, and each stage focuses on different contents. In China, software V&V on the NPP has been started from LingAo Phase II project and strongly done on other CPR1000 projects, which is a Generation II + pressurized water reactor. Based on the practice about YangJiang units 5 and 6 projects, combined with the relevant laws and regulation standards, this study summarizes the characteristics of the independent third party software V&V, analyzes the key points and methods of V&V activities in the software development process. As a result, it is also benefit to the design, operation and maintenance of safety I&C System as technical reference.

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Radiocarbon Releases At the Krško Nuclear Power Plant

Radiocarbon ◽

10.1017/s0033822200053261 ◽

1997 ◽

Vol 39 (3) ◽

pp. 285-292 ◽

Cited By ~ 10

Author(s):

Barbara Vokal ◽

Ivan Kobal

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Tree Rings ◽

Indoor Air ◽

Nuclear Power ◽

Monitoring Program ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Radioactivity Monitoring

Since 1991, radiocarbon analyses of exhaust air have been part of the regular radioactivity monitoring program at the Krško Nuclear Power Plant (NPP), a Westinghouse 632 MWe pressurized water reactor (PWR). Activity of CO2 and hydrocarbons has been identified; the former contributes ca. 43%. A normalized release of total 14C of 0.219 TBqGWe−1a−1 was obtained. Indoor air 14C concentrations in selected rooms inside the plant have generally been <5 Bq m−3, although rare peaks of >1000 Bq m−3 may be reached. Tree rings have shown slight enhanced 14C activity due to the operation of the plant.

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