Evaluation of the Impact of SAMG on the Level-2 PSA Results of a Pressurized Water Reactor

2006 ◽  
Vol 155 (1) ◽  
pp. 22-33 ◽  
Author(s):  
Yu-Chih Ko ◽  
Ching-Hui Wu ◽  
Min Lee
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Impact ◽  
Level 2

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

An Estimation of the Dynamic Buckling Load for the Spacer Grid of Pressurized Water Reactor Fuel Assembly

2006 ◽  
Vol 326-328 ◽  
pp. 1603-1606 ◽  
Author(s):  
Sang Youn Jeon ◽  
Young Shin Lee
Keyword(s):  
Fuel Assembly ◽  
Dynamic Buckling ◽  
Buckling Load ◽  
Estimation Methods ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Water Reactor ◽  
Spacer Grids ◽  
The Impact

This study contains an estimation of the dynamic buckling load for the spacer grid of fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the dynamic buckling loads of spacer grid. The dynamic impact tests and analyses were performed to evaluate the impact characteristics of the spacer grids and to predict the dynamic buckling load of the full size spacer grid. The estimation results were compared with the test results for the verification of the estimation methods.

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

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
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.

A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis

2021 ◽  
Author(s):  
Suubi Racheal ◽  
Yongkuo Liu ◽  
Miyombo Ernest ◽  
Abiodun Ayodeji
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Research Education ◽  
User Expertise ◽  
Scale Down ◽  
Water Reactors ◽  
The Impact ◽  
And Training

Abstract The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.

scholarly journals A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

2015 ◽  
Vol 2015 ◽  
pp. 1-14 ◽  
Author(s):  
Diego Mandelli ◽  
Steven Prescott ◽  
Curtis Smith ◽  
Andrea Alfonsi ◽  
Cristian Rabiti ◽  
...  
Keyword(s):  
System Analysis ◽  
Test Case ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Margins ◽  
Particle Hydrodynamics ◽  
Station Blackout ◽  
New Generation ◽  
The Impact

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

scholarly journals KEBUTUHAN SDM UJI TAK RUSAK UNTUK INSPEKSI PRESERVICE PADA PEMBANGUNAN PLTN PERTAMA DI INDONESIA

2012 ◽  
Vol 6 (1) ◽  
pp. 88
Author(s):  
Sri Nitiswati
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Level 3 ◽  
Level 1 ◽  
Service Data ◽  
Level 2

KEBUTUHAN SDM UJI TAK RUSAK UNTUK INSPEKSI PRE-SERVICE PADA PEMBANGUNAN PLTN PERTAMA DI INDONESIA. Peraturan Presiden Nomor 5 tahun 2006 tentang “kebijakan energi nasional” menyebutkan bahwa Indonesia telah memasukkan opsi nuklir ke dalam energi mix nasional untuk mendukung kebutuhan energi di masa mendatang. Artinya pembangunan pembangkit listrik tenaga nuklir (PLTN) sudah menjadi salah satu pilihan bangsa Indonesia untuk memenuhi kebutuhan energi. Pembangunan PLTN tidak bisa terlepas dari kebutuhan sumber daya manusia (SDM) termasuk SDM uji tak rusak yang penting pada tahap pabrikasi komponen PLTN yaitu untuk melakukan pemeriksaan/inspeksi komponen hasil fabrikasi atau yang biasa disebut dengan inspeksi pre-service. Data pemeriksaan komponen hasil dari fabrikasi digunakan sebagai informasi data awal kondisi komponen PLTN sebelum dioperasikan. Makalah ini membahas kebutuhan SDM uji tak rusak untuk inspeksi pre-service pada pembangunan PLTN pertama di Indonesia khususnya untuk memeriksa/inspeksi hasil pabrikasi komponen. Metode dengan cara melakukan identifikasi komponen-komponen yang berada di dalam sistem primer reaktor PWR (Pressurized Water Reactor), identifikasi bagian dari komponen yang harus diperiksa, metode uji tak rusak dan tekniknya serta kebutuhan SDM nya. Tujuannya adalah diperolehnya kebutuhan SDM uji tak rusak untuk melakukan inspeksi pre-service pada pembangunan PLTN I di Indonesia. Dari studi yang dilakukan disimpulkan bahwa SDM dengan kompetensi uji tak rusak metode visual, volumetrik dan permukaan diperlukan untuk inspeksi pre-service komponen-komponen hasil fabrikasi. Kebutuhan SDM berdasarkan volume pekerjaan adalah  untuk kompetensi UTR visual level 1: 10  orang, level 2 :8 orang, level 3 : 5 orang. Kompetensi UTR volumetrik level 1: 22 orang, level 2 : 13 orang, level 3 : 9 orang. Kompetensi UTR permukaan level 1: 8  orang, level 2 :7 orang, level 3 : 4  orang.

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