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Title: UNIC : ultimate neutronic investigation code.


No abstract prepared.

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Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
OSTI Identifier:
Report Number(s):
TRN: US201015%%1216
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: Joint International Topical Meeting on Mathematics & Computations and Supercomputing in Nuclear Applications (M&C + SNA 2007); Apr. 15, 2007 - Apr. 19, 2007; Monterey, CA
Country of Publication:
United States

Citation Formats

Palmiotti, G., Smith, M., Rabiti, C., Leclere, M., Kaushik, D., Siegel, A., Smith, B., Lewis, E.E., and Northwestern Univ. UNIC : ultimate neutronic investigation code.. United States: N. p., 2007. Web.
Palmiotti, G., Smith, M., Rabiti, C., Leclere, M., Kaushik, D., Siegel, A., Smith, B., Lewis, E.E., & Northwestern Univ. UNIC : ultimate neutronic investigation code.. United States.
Palmiotti, G., Smith, M., Rabiti, C., Leclere, M., Kaushik, D., Siegel, A., Smith, B., Lewis, E.E., and Northwestern Univ. Mon . "UNIC : ultimate neutronic investigation code.". United States. doi:.
title = {UNIC : ultimate neutronic investigation code.},
author = {Palmiotti, G. and Smith, M. and Rabiti, C. and Leclere, M. and Kaushik, D. and Siegel, A. and Smith, B. and Lewis, E.E. and Northwestern Univ.},
abstractNote = {No abstract prepared.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}

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  • No abstract prepared.
  • A code package called UNIC is currently under development at Argonne (Argonne National Laboratory). This new code package is focused on high fidelity solutions for nuclear reactor plant operations. The focus of this paper is the development of a spherical harmonics method used to solve the neutron transport equation. This new code, called P{sub N}FE, is just one of the components of UNIC. P{sub N}FE is targeted for use on massively parallel platforms using the PETSc library developed at ANL. In this paper, an overview of the theory behind the spherical harmonics method is given along with some results obtainedmore » with a parallel implementation of the code.« less
  • Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the mostmore » probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd= design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd. (authors)« less
  • Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions.