Evaluation of Diamond Detectors for Fast Neutron Fluence Measurements in WWER-1000 Surveillance Assemblies

2009 ◽  
pp. 467-467-8
Author(s):  
V. A. Nikolaenko ◽  
S. M. Zaritsky ◽  
I. V. Bachuchin
Keyword(s):  
Fast Neutron ◽  
Neutron Fluence

scholarly journals A Fast Neutron Fluence Reduction Strategy for Pressure Vessel Lifetime Extension

2000 ◽  
Vol 37 (sup1) ◽  
pp. 120-124 ◽  
Author(s):  
Jong Kyung Kim ◽  
Chang Ho Shin ◽  
Bo Kyun Seo ◽  
Myung Hyun Kim ◽  
Goung Jin Lee
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reduction Strategy ◽  
Lifetime Extension

scholarly journals A Method to Estimate the Fast-Neutron Fluence for the Hiroshima Atomic Bomb

1994 ◽  
Vol 63 (10) ◽  
pp. 3546-3547 ◽  
Author(s):  
Tokushi Shibata ◽  
Mineo Imamura ◽  
Seiichi Shibata ◽  
Yoshitomo Uwamino ◽  
Tohru Ohkubo ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Atomic Bomb ◽  
Neutron Fluence

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

Delayed Hydride Cracking Velocity of Irradiated Zr-2.5Nb Tubes after a 30-Year Operation in the Wolsong Unit 1

2005 ◽  
Vol 475-479 ◽  
pp. 1409-1414
Author(s):  
Young Suk Kim ◽  
Sang Bok Ahn ◽  
Kyung Soo Im ◽  
Wan H. Oh
Keyword(s):  
Yield Strength ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
Compact Tension ◽  
Round Robin ◽  
Velocity Dependence ◽  
Research Project ◽  
Round Robin Test ◽  
Delayed Hydride Cracking ◽  
Hydride Cracking

The aim of this study is to investigate a change in delayed hydride cracking (DHC) velocity of Zr-2.5Nb tubes with fast neutron fluence (E>1MeV) and predict the DHC velocity of the irradiated Wolsong 1 Zr-2.5Nb tubes at a neutron fluence corresponding to the 30 year design lifetime. To this end, the DHC velocity were determined at temperatures ranging from 100 to 280 oC on unirradiated Zr-2.5Nb tubes and the irradiated Zr-2.5Nb tubes in the Wolsong Unit-1 to the neutron fluence of 8.9x1025 n/m2 (E>1MeV). DHC tests were conducted on the compact tension specimens charged with 34 to 100 ppm hydrogen in accordance with the KAERI DHC procedures that have been validated through a round robin test on DHC velocity of Zr-2.5Nb tubes as an IAEA coordinated research project. Irradiated Zr-2.5Nb tubes had 3 to 5 times higher DHC velocity than that of unirradiated Zr-2.5Nb tubes while the inlet region of the irradiated Zr-2.5Nb tube with the highest yield strength had a slightly higher DHC velocity compared to that of the outlet region with the lowest yield strength. From a normalized correlation of yield strength and DHC velocity of the Zr-2.5Nb tubes, the yield strength was found to govern the DHC velocity of the Zr-2.5Nb tubes irrespective of the neutron fluence and operating temperatures. The DHC velocity of the irradiated Zr-2.5Nb tubes is predicted after a 30 year operation in the Wolsong Unit 1 on the basis of an increase in the yield strength with neutron fluence and a DHC velocity dependence on the yield strength of Zr-2.5Nb tubes.

The Feasibility Study of AGR 7-Position Fuel Testing Assembly in NEFT Position

2008 ◽  
Author(s):  
Gray S. Chang ◽  
Blaine Grover ◽  
John T. Maki ◽  
Misti A. Lillo
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Power Density ◽  
Neutron Spectrum ◽  
Neutron Fluence ◽  
Flux Ratio ◽  
Water Volume ◽  
Testing Time ◽  
Test Configuration ◽  
North East

In order to support the Next Generation Nuclear Plant (NGNP) Program 2018 deployment schedule, the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program must reduce the AGR fuel irradiation testing time in the Advanced Test Reactor (ATR) from approximately 2 1/2 calendar years to 1 1/2 calendar years. The AGR fuel irradiation testing requirements are: (a) burn-up of at least 14% FIMA; (b) Fast neutron fluence (E > 0.18 MeV) – maximum < 5.1 × 1025 n/m2; (c) limit of fission power density is 350 W/cc; and (d) irradiation time < 1 1/2 calendar years. The accelerated testing could be accomplished by utilizing the ATR North East flux trap (NEFT) position, which can provide more control of the thermal neutron flux rate than the ATR B-10 position currently being used for the AGR-1 fuel testing, which is regulated to achieve the fuel temperature and burn-up rate requirements. In addition, the Fast (E > 1.0 MeV) to Thermal (E < 0.625 eV) neutron flux ratio (F/T) for the NEFT is much harder (higher) than the F/T ratio for the B-10 position. Thus, an appropriate configuration of Beryllium (Be) and water will need to be determined in order to soften (lower) the F/T ratio to the desired value. The proposed AGR 7-position fuel test configuration in the NEFT will utilize a graphite holder consisting of six fuel specimen positions arranged around the perimeter of the graphite holder with a seventh fuel specimen position in the center of the holder. To soften the neutron spectrum in the fuel compacts, the water volume in the outer water annulus can be increased. To reduce the compact power density, a hafnium filter could be incorporated around the graphite holder. After several trials, a hafnium filter with a thickness of 0.008 inches appeared to adequately reduce the power density to achieve the fuel testing requirements. It was also determined that the chosen beryllium-tube and water annulus configuration would adequately soften the neutron spectrum to achieve the fuel testing requirement. This neutronics study is based upon typical ATR cycle operation of 50 effective full power days (EFPD) per cycle for seven proposed irradiation cycles, and a NE lobe power of the 14 MW. The MCWO-calculated fuel compact power density, burnup (% FIMA), and fast neutron fluence (E > 0.18 MeV) results indicate that the average fuel compact burnup and fast neutron fluence reach 14.79% FIMA and 4.16 × 1025 n/m2, respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 × 1025 n/m2, which satisfied the fuel testing requirements. It is therefore concluded that accelerating the AGR fuel testing using the proposed AGR 7-position fuel test configuration in the NEFT is very feasible.

Sign in / Sign up

Export Citation Format

Share Document