40. Three Mile Island reactor vessel investigation and findings

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

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Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2017/8294653 ◽

2017 ◽

Vol 2017 ◽

pp. 1-7

Author(s):

Mazhyn K. Skakov ◽

Nurzhan Ye. Mukhamedov ◽

Alexander D. Vurim ◽

Ilya I. Deryavko

Keyword(s):

Heat Capacity ◽

Thermal Diffusivity ◽

Fast Reactor ◽

Thermophysical Properties ◽

Heat Conductivity ◽

Nuclear Reactor ◽

Reactor Vessel ◽

Full Scale ◽

Temperature Fields ◽

First Time

For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel.

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Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

Journal of Physics Conference Series ◽

10.1088/1742-6596/962/1/012044 ◽

2018 ◽

Vol 962 ◽

pp. 012044 ◽

Cited By ~ 2

Author(s):

Amir Hamzah ◽

Suwoto ◽

Anis Rohanda ◽

Hery Adrial ◽

Syaiful Bakhri ◽

...

Keyword(s):

Neutron Fluence ◽

Reactor Vessel ◽

Pebble Bed ◽

Rate Analysis

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Effects of Delta - Ferrite Content on the Integrity of Reactor Vessel Cladding

ASME 2010 Pressure Vessels and Piping Conference: Volume 9 ◽

10.1115/pvp2010-25434 ◽

2010 ◽

Author(s):

Ho-Sang Shin ◽

Jin-Ki Hong ◽

Koo-Kab Chung ◽

Hae-Dong Chung ◽

Gwang-Yil Kim ◽

...

Keyword(s):

Electron Microscopy ◽

Scanning Electron Microscopy ◽

Reactor Vessel ◽

Bending Test ◽

Ferrite Content ◽

Delta Ferrite ◽

Bending Tests ◽

Asme Code ◽

Two Phases ◽

Scanning Electron

As the design life of new nuclear power plant increases, the austenitic stainless cladding integrity of reactor vessel becomes one of the new concerns. Since 1970’s, there have been some specific recommendations on delta ferrite content of austenitic cladding of reactor vessels and welds. It has been known that the delta ferrite is beneficial for reducing micro-fissure in welds, though the high delta ferrite content increases the probability of embrittlment of welds. In this study, the mechanical and microstructural properties of austenitic weld metals with the limit values of the recommended range (5 ∼ 18 FN) of the delta ferrite control on low alloy steels were characterized by using bending test and scanning electron microscopy. The base metal was ASME Code Sec. II specification SA 508 Gr. 3 Cl. 1 plate and weld materials were EQ308L and EQ309L strips. Four kinds of cladding were deposited with submerged arc welding process on SA508 cl.3 plates. The bending tests were performed through ASME code Sec. IX and the microstructure of fractured surfaces was analyzed by scanning electron microscopy (SEM). In bending tests, there were no fractures except the highest delta ferrite content specimens (28FN). From the SEM observation of fractured surfaces, cracks initiated from the interface between austenite and ferrites phases in the cladding layer and propagated through the continuous interfaces between two phases. For specimens without continuous interfaces of two phases, though the cracks were observed in the interface of phases, the propagation of cracks was not observed. From the test results, continuous interfaces between austenite matrix and ferrite phase provide the path for crack propagation. And the delta ferrite content affects the integrity of cladding of reactor vessel.

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Study on the Heat Transfer in the Water Pool Type Reactor Cavity Cooling System of the Very High Temperature Gas Cooled Reactor

Heat Transfer: Volume 4 ◽

10.1115/ht2005-72589 ◽

2005 ◽

Author(s):

H. K. Cho ◽

D. U. Seo ◽

M. O. Kim ◽

G. C. Park

Keyword(s):

Heat Transfer ◽

Forced Convection ◽

Cooling System ◽

Reactor Vessel ◽

Water Pool ◽

Cooling Pipe ◽

Cavity Cooling ◽

The Heat Transfer Coefficient ◽

Pool Type ◽

High Temperature Gas

In the HTGR (High Temperature Gas Cooled Reactor), the Reactor Cavity Cooling System (RCCS) is equipped to remove the heat transferred from the reactor vessel to the structure of the containment. The function of the RCCS is to dissipate the heat from the reactor cavity during normal operation including shutdown. The system also removes the decay heat during the loss of forced convection (LOFC) accident. A new concept of the water pool type RCCS was proposed at Seoul National University. The system mainly consists of two parts, water pool located between the containment and reactor vessel and five trains of air cooling system installed in the water pool. In normal operations, the heat loss from the reactor vessel is transferred into the water pool via cavity and it is removed by the forced convection of air flowing through the cooling pipes. During the LOFC accident, the after heat is passively removed by the water tank without the forced convection of air and the RCCS water pool is designed to provide sufficient passive cooling capacity of the after heat removal for three days. In the present study, experiments and numerical calculations using CFX5.7 for the water pool and cooling pipe were performed to investigate the heat transfer characteristics and evaluate the heat transfer coefficient model of the MARS-GCR (Multi-dimensional Analysis of Reactor Safety for Gas Cooled Reactor Analysis) which was developed for the safety analysis of the gas cooled reactor. From the results of the experiments and CFX calculations, heat transfer coefficients inside the cooling pipe were calculated and those were used for the assessment for the heat transfer coefficient model of the MARS-GCR.

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An Improved Test Facility for Studying Deposit Fouling on Steam Turbine Blades

Volume 8: Microturbines, Turbochargers and Small Turbomachines; Steam Turbines ◽

10.1115/gt2016-57575 ◽

2016 ◽

Author(s):

Alan R. May Estebaranz ◽

Richard J. Williams ◽

Simon I. Hogg ◽

Philip W. Dyer

Keyword(s):

Turbine Blades ◽

Steam Pressure ◽

Reactor Vessel ◽

Test Facility ◽

Test Results ◽

Steam Turbines ◽

Steam Power ◽

Scale Test ◽

Steam Turbine Blades ◽

Asme Paper

A laboratory scale test facility has been developed to investigate deposition in steam turbines under conditions that are representative of those in steam power generation cycles. The facility is an advanced two-reactor vessel test arrangement, which is a more flexible and more accurately controllable refinement to the single reactor vessel test arrangement described previously in ASME Paper No. GT2014-25517 [1]. The commissioning of the new test facility is described in this paper, together with the results from a series of tests over a range of steam conditions, which show the effect of steam conditions (particularly steam pressure) on the amount and type of deposits obtained. Comparisons are made between the test results and feedback/experience of copper fouling in real machines.

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Mechanical Stress Improvement Process (MSIP) Used to Prevent and Mitigate Primary Water Stress Corrosion Cracking (PWSCC) in Reactor Vessel Piping at V.C. Summer

Aging Management and Component Analysis ◽

10.1115/pvp2003-2160 ◽

2003 ◽

Author(s):

E. A. Ray ◽

K. Weir ◽

C. Rice ◽

T. Damico

Keyword(s):

Water Stress ◽

Stress Corrosion ◽

Mechanical Stress ◽

Stress Corrosion Cracking ◽

Reactor Vessel ◽

Corrosion Cracking ◽

The Other ◽

Water Reactor ◽

Improvement Process ◽

Primary Water

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.

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