Investigation of pressure vessel neutron fluence calculation with Monte Carlo

Conference · Fri Jan 01 00:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5884135

Wagner, J C; Haghighat, A; Petrovic, B G ^[1]

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1. Pennsylvania State Univ., University Park (United States)

The life of a reactor and its possible extension are strongly dependent on the embrittlement of the pressure vessel under neutron irradiation. Therefore, in order to assess pressure vessel radiation damage, the neutron fluence must be accurately determined. The S[sub N] transport method, is currently used to determine a synthetic three-dimensional flux distribution based on one- and two-dimensional transport calculations. In the past, these calculations have been benchmarked based on data from in-vessel capsules, but due to mechanical problems and subsequent safety issues associated with these capsules, ex-vessel cavity dosimetry is being considered as an alternative. Additional uncertainties, however, may be introduced into the neutron transport calculational results because of the presence of an optically thick region of stainless steel (PV) and three-dimensional streaming through the void (cavity region). Hence, it is necessary to evaluate these uncertainties with the Monte Carlo method. The two principal advantages of the Monte Carlo method are explicit geometric representation and continuous-energy simulation. This paper presents an MCNP mode 12 for determining pressure vessel neutron fluence and compares the Monte Carlo results with the deterministic and experimental results for Three Mile Island unit 1 (TMI-1) cycle 7.

• OSTI ID: 5884135
• Report Number(s): CONF-930601-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 68; Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
36 MATERIALS SCIENCE
NEUTRON TRANSPORT
COMPUTERIZED SIMULATION
PRESSURE VESSELS
EMBRITTLEMENT
REACTOR VESSELS
DAMAGE
IRRADIATION
MONTE CARLO METHOD
NEUTRON FLUENCE
OPTIMIZATION
THREE-DIMENSIONAL CALCULATIONS
CALCULATION METHODS
CONTAINERS
NEUTRAL-PARTICLE TRANSPORT
RADIATION TRANSPORT
SIMULATION
220100* - Nuclear Reactor Technology- Theory & Calculation
663610 - Neutron Physics- (1992-)
360106 - Metals & Alloys- Radiation Effects