Improved Reactor Vessel and Bottom Nozzle UT Inspection Experience

2005 ◽  
Author(s):  
S. W. Glass ◽  
A. Richmond ◽  
G. Alexander
Keyword(s):  
Pressure Vessel ◽  
Field Experience ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Inspection Period ◽  
New Techniques ◽  
Nuclear Station ◽  
Service Inspection ◽  
Reactor Pressure

Framatome ANP, an AREVA and Siemens company, recently completed a combined ten-year reactor pressure vessel (RPV) in-service inspection (ISI) and a volumetric ultrasonic (UT) examination of all bottom-mounted nozzles (BMNs) at Florida Power & Light’s (FPL) Turkey Point-3 nuclear station. Both the 10-year RPV robot and the BMN tool had undergone significant improvements since their last deployment. The enhancements focused primarily on reducing inspection times. The new tools were also designed to minimize the in-containment lay-down area and utility support during the inspection period. This latter aspect was particularly critical for FPL during this outage because of the busy refuel floor schedule that included both vessel examination and head replacement tasks. It was logical to combine these two examinations since they are both performed with the internals completely removed. The vessel examination required approximately 4 days (95.5 hours) and the BMN examination required 1.5 days (36 hours). Total “hook-to-hook” time was 5.75 days (138 hours) — more than 1 day ahead of schedule. The dose savings over alternate inspection approaches for the BMN examination was also very significant. This paper discusses the tool improvements and new techniques used for the examinations of the RPV and the BMNs as well as the field experience during the FPL examination.

Design and Fabrication of Reactor Pressure Vessel for High Temperature Engineering Test Reactor (HTTR)

2004 ◽  
Author(s):  
Yukio Tachibana ◽  
Shigeaki Nakagawa ◽  
Tatsuo Iyoku
Keyword(s):  
Pressure Vessel ◽  
Temper Embrittlement ◽  
Reactor Pressure Vessel ◽  
The Body ◽  
Inlet Temperature ◽  
Irradiation Embrittlement ◽  
Fabrication Procedure ◽  
Test Reactor ◽  
Service Inspection ◽  
Reactor Pressure

The reactor pressure vessel (RPV) of the HTTR is 5.5 m in inside diameter, 13.2 m in inside height, and 122 mm and 160 mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1×1017 n/cm2 (E>1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.

Reactor Pressure Vessel of Tricastin Unit 1 Core Zone Inspection: Comparison of Non-Destructive Examination Results Among Two Core Area Inspections Performed in 1999 and in 2003

2004 ◽  
Author(s):  
Karl Payraudeau ◽  
Karim Zamoum ◽  
Thierry Pasquier
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Zone ◽  
Ultrasonic Process ◽  
Dimension Change ◽  
First Time ◽  
Non Destructive ◽  
Service Inspection ◽  
Reactor Pressure

As part of the unit aging follow-up, a non.destructive examination has been designed to inspect the core zone of the Reactor Pressure Vessel (RPV). This ultrasonic process called “VPM” has been developed and qualified by Intercontroˆle in accordance with EDF specifications. The qualification has been attested by an accredited qualification body. In 1999, the VPM process was used the first time during the in-service inspection of TRICASTIN, unit 1. As a result of the RPV core zone inspection a set of under cladding flaw indications was detected and characterized with the VPM process. These flaws have been analysed as planar manufacturing flaws. In 2003, 4 cycles later, during the TRICASTIN Unit 1 outage inspection, “VPM” was used to characterize again this set of flaw indications in order to verify their dimensions. The changes of the main characteristics (specially height) between both the inspections were compared to the process accuracy. No significant dimension change has been recorded.

RELAP Simulation on Boiling and Flow Phenomena Under External Reactor Vessel Cooling Condition

2010 ◽  
Author(s):  
Yongchun Li ◽  
Weihua Zhou ◽  
Yanhua Yang ◽  
Bo Kuang ◽  
Xu Cheng
Keyword(s):  
Pressure Vessel ◽  
Flow Boiling ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Study Program ◽  
External Cooling ◽  
Flow Phenomena ◽  
Key Issues ◽  
Reactor Pressure

External reactor vessel cooling (ERVC) of the In-vessel retention (IVR) system is widely accepted as a feasible way to remove decay heat from the lower head of the reactor pressure vessel (RPV) under severe accident (SA) conditions. However, some issues relating to ERVC still need to be evaluated before its application, such as boiling and flow phenomena and CHF prediction, etc. To study these key issues, an experimental study program named REPEC (Reactor Pressure Vessel External Cooling) is performed at Shanghai Jiao Tong University. Steady state experiments focusing on flow boiling phenomena investigation are carried out with comprehensive measurements, including temperature distribution, pressure drop and mass flow rate. As a part of studies on boiling mechanism and flow phenomena between RPV and the insulation, the experiment is analyzed and simulated with RELAP code. The code simulation covers most of the experimental cases, and a comparison between simulation results and experimental data are presented and discussed.

Monitoring of a Radiation Embrittlement by a Reliable Fluence Evaluation With a Neutron Dosimetry in a Reactor Pressure Vessel

2006 ◽  
Author(s):  
SamLai Lee ◽  
ByoungChul Kim ◽  
ChoonSung Yoo ◽  
KeeOk Chang
Keyword(s):  
Pressure Vessel ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Transport ◽  
Axial Direction ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Neutron Dosimetry ◽  
Good Reliability ◽  
Reactor Pressure

The assurance of a nuclear reactor pressure vessel integrity plays an important role in achieving a safety and extending the life of nuclear power plants. In the assessment of the state of an embrittlement of a pressure vessel in a pressurized light water reactor, an accurate evaluation of the neutron exposure of each of the materials comprising the beltline region of the vessel is required. The evaluation is performed through a neutron dosimetry analysis where a fluence calculation is done by both measurement of the dosimeter materials removed from surveillance capsules and a neutron transport calculation. Now that all the capsules have been completely removed from the reactor vessel and analyzed by a periodic monitoring schedule, four ex-vessel sensor sets are installed as a substitute capsule in an axial direction in the reactor cavity and then removed for an analysis in order to meet the regulation requirements. The results showed that the differences between the measurements and calculations are less than 20% for each capsule, which means these analyses satisfy the acceptable criterion required by Regulatory Guide 1.190, and they also provide an assurance that such an evaluation including an ex-vessel neutron dosimetry can be used to predict the fluence of a nuclear reactor vessel with a good reliability.

Simulation of Reactor Pressure Vessel Internals Thermal Stress Due to Internal Heat Generation and Environmental Boundary Conditions Using Stress Transfer Functions

2008 ◽  
Author(s):  
Mark A. Gray ◽  
E. Lyles Cranford ◽  
Charles B. Gilmore ◽  
Ste´phane G. Guillot ◽  
Chuang Y. Yang
Keyword(s):  
Heat Generation ◽  
Pressure Vessel ◽  
Thermal Stresses ◽  
Transfer Functions ◽  
Stress Transfer ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Internal Structures ◽  
Fluid Transients ◽  
Reactor Pressure

The internal heating rates from the capture and scatter of neutrons and primary gamma radiation represent one of the most significant loadings acting on the reactor pressure vessel internal structures. The nuclear heating effects on the internal structures are mitigated by the flow of the primary coolant removing heat energy from these structures. This paper discusses the use of transfer functions to predict the time-dependent thermal stresses in the lower core plate of reactor vessel internals. Numerous stress analysis applications have employed the transfer function approach to obtain thermal stresses from fluid transients in reactor pressure vessel and piping locations. These solutions have been utilized in both design analysis applications and in stress and fatigue monitoring roles for evaluation of pressure vessels subjected to fluid transients and mechanical loads such as pressure and piping loads. This paper presents a further extension of the technology to address heat generation loads. The evaluation of stress and fatigue produced in a reactor vessel internals lower core plate due to multiple loadings, including heat generation, using the stress transfer function approach was investigated for reducing analysis process time and conservatism when compared to conventional analysis methods. Reduced conservatism provides margin that may be beneficial in life extension studies.

scholarly journals Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Alejandro Nuñez-Carrera ◽  
Raúl Camargo-Camargo ◽  
Gilberto Espinosa-Paredes ◽  
Adrián López-García
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Cooling Water ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Detailed Model ◽  
Lower Head ◽  
Reactor Pressure

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

Severe Accident Improvements for CAREM-25 to Arrest Reactor Vessel Meltdown Sequences

2002 ◽  
Author(s):  
L. E. Pomier Ba´ez ◽  
J. E. Nun˜ez Mac Leod ◽  
J. H. Baro´n
Keyword(s):  
Power Plant ◽  
Pressure Vessel ◽  
Transient Behavior ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Element Analysis ◽  
Management Actions ◽  
Core Catcher ◽  
Reactor Pressure

Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the analyses, via numerical simulation, for the conceptual design of such a container type. Simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features is addressed. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. The finite element analysis (FEA) system ALGOR has been used to evaluate the thermal regime of the reactor pressure vessel wall, when the in-vessel metallic core catcher is present and when it is not present. Two different scenarios have been considered for heat transfer outside the reactor vessel, a pessimistic (dry) and optimistic (wet) conditions in the reactor cavity. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher.

Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

2002 ◽  
Author(s):  
Allen L. Hiser
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Reactor Pressure Vessel ◽  
Regulatory Requirements ◽  
Pressurized Water ◽  
Head Penetration ◽  
Nuclear Station ◽  
Station Unit ◽  
Reactor Pressure

The recent discoveries of cracked and leaking Alloy 600 vessel head penetration (VHP) nozzles, including control rod drive mechanism (CRDM) and thermocouple nozzles, at four pressurized water reactors (PWRs) have raised concerns about the structural integrity of VHP nozzles throughout the PWR industry. Nozzle cracking at Oconee Nuclear Station Unit 1 in November 2000 and Arkansas Nuclear One Unit 1 in February 2001 was limited to axial cracking, an occurrence deemed to be of limited safety concern in the NRC staff’s generic safety evaluation on the cracking of VHP nozzles dated November 19, 1993. However, the discovery of circumferential cracking at Oconee Nuclear Station Unit 3 in February 2001 and Oconee Nuclear Station Unit 2 in April 2001 particularly the large circumferential cracking identified in two CRDM nozzles at ONS3 has raised concerns about the potential safety implications and prevalence of cracking in VHP nozzles in PWRs. In response to the circumferential cracking identified at the Oconee units, the NRC issued Bulletin 2001–01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” on August 3, 2001. This bulletin requests information from licensees related to the structural integrity of the reactor pressure VHP nozzles for their facilities, including the extent of VHP nozzle leakage and cracking that has been found to date, the inspections and repairs that have been undertaken to satisfy applicable regulatory requirements, future plans to inspect VHP nozzles, and a description of how future inspection plans will ensure compliance with applicable regulatory requirements. This paper summarizes the staff’s review and assessment of licensee responses to NRC Bulletin 2001–01.

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