Analysis of Thermal-Hydraulic Fluctuations in Trillo NPP With CTF/PARCS v. 2.7 Coupled Code

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
A. Abarca ◽  
R. Miró ◽  
G. Verdú ◽  
J. A. Bermejo
Keyword(s):  
Reactor Core ◽  
Low Frequency ◽  
Protection System ◽  
Inlet Temperature ◽  
Frequency Noise ◽  
Pressurized Water ◽  
The Core ◽  
Coupled Codes ◽  
Water Reactors ◽  
New Research

The low-frequency noises are fluctuations in the neutron flux density, in the low-frequency range up to 4 Hz, which generate noise in the neutron instrumentation and could affect the limitation and protection system of the reactor core. Some European pressurized water reactors (PWRs) experienced the effect of low-frequency noise, opening a new research line for the verification of the neutron-kinetics/thermal-hydraulic coupled codes. A CTF/PARCS v. 2.7 simulation study to verify whether periodical fluctuations in the core inlet temperature could activate the core protection system has been done, obtaining the frequency spectrum of the power oscillation amplitudes.

scholarly journals SENSITIVITY ANALYSIS OF STAINLESS STEEL REFLECTOR FOR VR-1 TRAINING REACTOR

2021 ◽  
Vol 247 ◽  
pp. 08003
Author(s):  
Jan Frybort ◽  
Lubomir Sklenka ◽  
Filip Fejt ◽  
Pavel Suk ◽  
Lenka Frybortova
Keyword(s):  
Stainless Steel ◽  
Power Distribution ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Fast Neutrons ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Radial Reflector

Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics. The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors’ impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Irradiation Hardening of Austenitic Stainless Steels: From Field Experience to Modeling

2004 ◽  
Author(s):  
C. Pokor ◽  
Y. Thebault ◽  
C. Pujol ◽  
J.-P. Massoud ◽  
D. Loisnard ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactors ◽  
Iterative Approach ◽  
Pressurized Water ◽  
Irradiation Hardening ◽  
The Core ◽  
Water Reactors ◽  
Life Predictions

Materials for the core internals of Pressurized Water Reactors (austenitic stainless steels) are submitted to neutron irradiation. To understand the ageing mechanisms associated to irradiation and propose life predictions of the component, a multi step iterative approach consisting in particular in modeling the evolution of the hardening has been undertaken. Combination of characterization and modeling of simplified situations and field expertise is proposed.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

Leisure Places Downstairs: A Pleasure or a Failure?

1998 ◽  
Vol 5 (2) ◽  
pp. 79-90
Author(s):  
M. Asselineau
Keyword(s):  
Urban Planning ◽  
Urban Environment ◽  
Low Frequency ◽  
Frequency Noise ◽  
Low Frequency Noise ◽  
Community Noise ◽  
The Core ◽  
European Cities ◽  
New Buildings

One of the urban environment challenges is to manage to locate such leisure places as cinemas, restaurants, or even discotheques, as close as possible to living quarters. While this can be reasonably achieved in brand new buildings, with careful acoustical and urban planning and engineering, it often proves tricky, or even impossible, to achieve in the kind of older buildings that usually are to be found at the core of European cities. Whenever any benefit results from the presence of such leisure places close to homes, the neighbourhood can much more readily accept the acoustical implications. However, when no thought is given to the acoustical problems, the technical and relational efforts needed to correct the situation often prove to be beyond the capabilities of the operators. Over the years, the authorities have tried to brush up community noise regulations so that they can cope with the new trends that include low frequency noise involving music.

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

2020 ◽  
Vol 7 (3) ◽  
Author(s):  
Andrea Bachrata ◽  
Fréderic Bertrand ◽  
Nathalie Marie ◽  
Fréderic Serre
Keyword(s):  
Mitigation Strategies ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Physical Parameters ◽  
Fast Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Sodium Fast Reactors ◽  
Water Reactors ◽  
Accident Scenarios

Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

scholarly journals ICONE19-43115 ASSESSMENT OF ADVANCED REACTOR CORE PROTECTION SYSTEM FOR PRESSURIZED WATER REACTOR

2011 ◽  
Vol 2011.19 (0) ◽  
pp. _ICONE1943-_ICONE1943
Author(s):  
Wang-Kee In ◽  
Young-Ho Park ◽  
Chang-Ho Kim ◽  
Seung-Yeob Baeg ◽  
Tae-Young Yoon
Keyword(s):  
Reactor Core ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor
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