Hızlandırıcı güdümlü toryum reaktöründe serpent Monte Carlo kodu kullanılarak yanma oranı hesaplamaları

Yıl 2012, Cilt: 28 Sayı: 4, 309 - 314, 01.08.2012

Mehmet Emin Korkmaz Osman Ağar Mustafa Yiğit

Öz

Bu çalışma, Serpent Monte Carlo kodu kullanılarak Hızlandırıcı Güdümlü Toryum Reaktör’ünün (ADTR) yanma oranı (burnup) karakteristiğini elde etmek için yapılmıştır. Serpent Monte Carlo kodu reaktör örgü fizik hesaplamaları için optimize edilmiştir. ADTR modeli, merkezinde doğal kurşun - bizmut bir silindirik hedef, kor bölgesinde hegzagonal yapıda 90 yakıt demeti ve her demette 91 yakıt çubuğunda n oluşmuştur. Sistemin merkezinde bulunan hedefin uzunluğu 60 cm ve yarıçapı 15 cm’dir. Yakıt çubuklarının dağılımı 232 Th ve 233 U’un bulunduğu iki tip yakıt malzemesinden oluşmaktadır. Hesaplamalarda ENDF/B - 6.8, ENDF - 7, JEF - 2.2, JEFF - 3.1 değerlendirilmiş n ükleer datası ve yanma oranı hesaplamalarında “Chebyshev Rasyonel Yaklaşım Metodu” kullanılmıştır. Tasarlanan sistemde ortalama güç yoğunluğu ise 38.6E - 3 kW/g’dır. Kullanılan farklı nükleer tesir kesiti kütüphaneleri arasındaki değişiklikler incelenmiş ve uyumlu olduğu görülmüştür. Bütün yanma oranı hesaplamaları 40 MW/kgU’a kadar 41 farklı basamakta yapılmıştır.

Anahtar Kelimeler

Monte Carlo Yöntemi, Toryum, Yanma oranı, Kritikaltı reaktör

Burnup calculations for accelerator driven thorium reactor using serpent Monte Carlo code

Öz

In this work, a number of simulations have been performed in order to obtain neutronic and burnup characteristics of Accelerator Driven Thorium Reactor (ADTR) using Serpent Monte Carlo Code. This code is optimized for reactor lattice calculations. Center of ADTR model has a cylindrical natural Pb - Bi target, has 90 hexagonal structure fuel bundle in core and each bundle consist of 91 fuel pin. The length and radius of the target in the center of the system are 60 cm and 15 cm, respectively. The distribution of fuel rods consist of two types of fuel material with 232 Th and 233 U. The burnup calculat ions have been carried out using matrix exponential solution based on the Chebyshev Rational Approximation Method (CRAM) and the evaluated nuclear data used in the calculations was based on ENDF/B - 6.8, ENDF - 7, JEF - 2.2, JEFF - 3.1. The average power density i s 38.6E - 3 kW/g in designed system. Changes between the different nuclear cross section libraries were examined and observed to be compatible. All burnup calculations made up of 41 different steps up to 40 MW / kgU.

Anahtar Kelimeler

Monte Carlo Method, Thorium, Burnup, Subcritical Reactor

Kaynakça

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