scholarly journals A Criticality Analysis of the GBC-32 Dry Storage Cask with Hanbit Nuclear Power Plant Unit 3 Fuel Assemblies from the Viewpoint of Burnup Credit

2016 ◽  
Vol 48 (3) ◽  
pp. 624-634 ◽  
Author(s):  
Hyungju Yun ◽  
Do-Yeon Kim ◽  
Kwangheon Park ◽  
Ser Gi Hong
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Criticality Analysis ◽  
Nuclear Power Plant Unit

Model of a Dry Storage Facility for a Medium Nuclear Power Plant

2001 ◽  
Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Dose Rate ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Radiation Dose Rate ◽  
Dry Storage ◽  
Fuel Assemblies

Abstract The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

2021 ◽  
Author(s):  
Davor Grgic ◽  
Mario Matijevic ◽  
Paulina Duckic ◽  
Radomir Jecmenica
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Variance Reduction ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
The Impact

Abstract In this paper shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW loaded with Spent Fuel Assemblies (SFA) from Nuclear Power Plant (NPP) Krsko Spent Fuel Dry Storage (SFDS) Campaign one. The HI-TRAC VW is a multi-layered cylindrical vessel designed to accept a Multi Purpose Canister (MPC) during loading, unloading and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using ORIGEN-S module of the SCALE code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron induced gammas (n-g reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, FW-CADIS variance reduction scheme, as implemented in ADVANTG code, was adopted for accelerating final MCNP6 calculations. The dose rates around HI-TRAC VW cask were calculated using MCNP6 code for all 16 casks loading belonging to Campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization (Holtec International).

Criticality analysis for safety-critical software in nuclear power plant distributed control system

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 343-352
Author(s):  
J. Cui ◽  
Y. Cai ◽  
Y. Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Application Process ◽  
Software Module ◽  
Critical System ◽  
Safety Critical ◽  
Plant Safety ◽  
Safety Integrity Level ◽  
Criticality Analysis

Abstract Software criticality analysis examines the degree of contribution that each individual failure mode of a software component has on the reliability of software. Higher safety integrity levels are assigned to software modules whose failures cause an unacceptable impact on the operation of the system, and these levels require the implementation of more rigorous software quality assurance measures as defined in IEEE Std 1012 and in the customer’s system requirements specification. In this paper, a novel software criticality analysis method is proposed, the results of which can be used to guide the development of newly developed software and the procurement of Commercial-Off-The-Shelf (COTS) software. The software structure is first analyzed and the software is divided into modules according to their functions. Then the criticality levels of software components are preliminarily classified by means of a safety criticality preliminary analysis tree, followed by their verification through the software hazard and operability analysis (HAZOP). Finally, the target Safety Integrity Level (SIL) of each software module is determined based on its criticality level and the overall safety objective (i. e., SIL) of the system it resides in. As an example, this proposed method is applied to a nuclear power plant safety-critical system to demonstrate the detail application process and to verify the feasibility of the method. Compared with the existing software criticality analysis methods, this method has better operability and verifiability, and can be utilized as a technical guidance for the software criticality analysis of nuclear power plant digital control systems.

Risk Based Inspection Pilot Study of Ignalina Nuclear Power Plant, Unit 2

2002 ◽  
Author(s):  
Bjorn Brickstad ◽  
Adam Letzter ◽  
Arturas Klimasauskas ◽  
Robertas Alzbutas ◽  
Linas Nedzinskas ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Pilot Study ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Damage Mechanism ◽  
Core Damage ◽  
Steel Welds ◽  
Nuclear Power Plant Unit

A project with the acronym IRBIS (Ignalina Risk Based Inspection pilot Study) has been performed with the objective to perform a quantitative risk analysis of a total of 1240 stainless steel welds in Ignalina Nuclear Power Plant, unit 2 (INPP-2). The damage mechanism is IGSCC and the failure probabilities are quantified by using probabilistic fracture mechanics. The conditional core damage probabilities are taken from the plant PSA.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

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