Reactor Physics Quarterly Report (July, August, September 1969)

Technical Report · Sat Nov 01 04:00:00 EST 1969

DOI:https://doi.org/10.2172/4742013· OSTI ID:4742013

Schmid, L. C. ^[1]; Clayton, E. D. ^[1]; Heineman, R. E. ^[1]

---

1. Battelle Pacific Northwest Labs., Richland, WA (United States)

REACTOR THEORY AND CODE DEVELOPMENT: Battelle-Northwest Program Gas, a gas loop analysis program, has been modified to treat gases with temperature dependent specific heats. A general method for imposing explicit equality constraints on the functional parameters in the nonlinear least-squares program LEARN-I-A has been developed and demonstrated. Major changes have been made to the least-squares code, which is used to extract one group cross section information from isotopic concentration burnup data. The resulting new programs are called BUFIT (burnup fitting code) and a double precision version DBUFIT. The definition of the infinite multiplication factor has been clarified and related to the parameter usually calculated by theory. In addition, leakage terms required for using an unpoisoned technique have been derived. THERMAL REACTORS: The fourth interim critical tests were conducted on the 55 fuel element Batch Core loading at the PRTR at an average core burnup of 5330 MWd/MTM. Tests were also conducted on two smaller loadings of 31 and 19 fuel elements. Destructive burnup data from samples cut from PRTR rods have been evaluated for consistency and processed through the ISODIL burnup analysis code. Calculations of the fast effect for uranium fueled-H₂O lattices have shown that the error incurred in assuming a homogeneous cell is less than 0.5% in k_eff. Thus, if all other sources of error were eliminated in the calculations, values of k_eff 5 1.005 should be calculated for the UO₂ and the U02-PuO₂ fueled water lattices. An analytical study was made to determine the sensitivity of calculated values of k_eff to the amount of energy and angular detail used in a transport theory solution for a thermal reactor problem. The results show that a 10-group S-8 solution is accurate to within +/- 0.2% in k_eff. An equivalence approximation has been developed for use in applying the Dunn flux peaking or dipping correction to pin activations. A comparison was made between the results of burnup calculations obtained using the ZODIAC G and ALTHAEA codes. As a result of this comparison, it is concluded that the ALTHAEA code can be used in studies of reactor operation and behavior for plutonium fueled cores as well as uranium fueled cores. FAST REACTORS: Neutron spectrum studies have been made in the Physical Constants Test Reactor Fast Neutron Cavity using both a proton recoil spectrometer and a multiple foil activation method. Comparison of the two experimental methods shows that information obtained from each supplements and confirms the results of the other. Composite results of these measurements provide a valuable check of both reactor theory and analytical technique for developing calculational methods for fast- thermal coupled reactor cores. Comparisons of the group cross sections generated from 2 three codes (ETOX-lDX, PIC², and GAFGAR) were compared for the Fast Reactor ZPR III Assembly 52. The Beam Bending Code Development Program has progressed to the point where major decisions are being made for the model formulation and code computing procedures. The code to be developed will be a modification of AXISOL-AXICRP. This general finite element procedure will permit the creep bending analysis of the flow ducts using the same structural idealization as used for elastic bending analyses. A matrix decomposition scheme that has been developed for the code's application simplifies solution of the equations governing the beam behavior. Differential swelling in the fuel assemblies result in duct bending which is the major fuel lifetime -limiting item. Additionally, fuel pin packing within the duct may occur between 30 and 80,000 MWd/MTM, depending upon the metal swelling models used. CRITICAL MASS PHYSICS: The HAMMER system has been found to calculate, with reasonable accuracy, material bucklings measured experimentally for 1.5, 2.0, 4.0 and 6.6 wt% PuO₂/PuO₂₂ rods in water. Values of reflector savings are presented for uranyl nitrate of low enrichment (<5 wt% ²³⁵U) reflecting a spherical lattice of uranium rods in water. Curves of reflector savings as a function of k_∞ for a reflector of fixed thickness are nearly independent of the ²³⁵U enrichment. Calculations were performed by HFN computer program.

• Research Organization: Battelle Pacific Northwest Labs., Richland, WA (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
• DOE Contract Number: AT(45-1)-1830
• NSA Number: NSA-24-007168
• OSTI ID: 4742013
• Report Number(s): BNWL-1240
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ACCIDENTS
BEHAVIOR
BURNUP
COMPUTERS
CRITICALITY
Code Development
Critical Mass Physics
Criticality
DESIGN
DIGITAL SYSTEMS
Dynamics
EXPANSION
FAST NEUTRONS
FAST REACTORS
FTR
FUEL ELEMENTS
Fast Reactors
Kinetics
N38510* --Research &amp; Test Reactors &amp; Critical Assemblies-- Kinetics &amp; Dynamics
N38910 --Reactor Theory--Kinetics
NEUTRON FLUX
Nuclear Criticality Safety Program (NCSP)
PCTR
PLUTONIUM OXIDES
PROGRAMMING
PRTR
REACTOR CORE
RESEARCH REACTORS
Reactor Theory
SPECTRA
SWELLING
THERMAL NEUTRONS
TRANSPORT THEORY
Thermal Reactors
URANIUM DIOXIDE