Search Menu
DOE PAGES title logo U.S. Department of Energy
Office of Scientific and Technical Information
Search Scholarly Publications

Title: Upgrading Limiting Peak-Power Analysis Techniques with Modern Validation and Uncertainty Quantification for the Advanced Test Reactor

Journal Article · · Nuclear Technology
ORCiD logo [1];  [1];  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
+ Show Author Affiliations

Here, this work demonstrates the acceptability of the 2D deterministic transport code, HELIOS, to replace the legacy diffusion code, PDQ, for computing the peak-power performance parameters of the Advanced Test Reactor (ATR). The 95% Confidence Rule, commonly used in the commercial reactor sector, is explored to develop the so-called “reliability factors” which provide statistical confidence that the peak-power limits within the hottest location along a fuel plate, referred to as the hot-stripe, will not be exceeded. Additionally, an alternative “legacy” methodology was explored that attempts to mimic the exact PDQ analysis process used for defining the peak-power limits. The legacy methodology, involves interpolating power between regions at azimuthal boundaries subtending the regions of interest. In order to apply the 95% Confidence Rule, a statistically significant calculation-to-measurement bias must first be established. Unlike the commercial world where thousands of power observations can be collected every cycle using on-line flux mapping instrumentation, the ATR power distribution must be measured during “depressurized” zero-power measurements using fission wires in polyethylene wands. In 2012, fission wire activation data was collected during a flux run in the Advanced Test Reactor – Critical facility. Also to improve statistical validity, archival data from ATR zero power flux runs from 1977, 1986, and 1994 were digitized from scanned reports and used to create new benchmark models. Borrowing from least-squares adjustment methods commonly used for neutron activation spectroscopy, adjusted fission wire powers were calculated for all four datasets. The mean and standard deviation of the bias between a priori calculated and adjusted wire-powers was then taken as the bias and uncertainty used in the 95% Confidence Rule.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
1473711
Report Number(s):
INL/JOU-17-41057-Rev000
Journal Information:
Nuclear Technology, Vol. 201, Issue 3; ISSN 0029-5450
Publisher:
Taylor & Francis - formerly American Nuclear Society (ANS)Copyright Statement
Country of Publication:
United States
Language:
English

References (8)

On measuring skewness and kurtosis journal September 2007
A fission matrix based validation protocol for computed power distributions in the advanced test reactor journal December 2015
An Extension of Shapiro and Wilk's W Test for Normality to Large Samples journal January 1982
An analysis of variance test for normality (complete samples) journal December 1965
Experimental Statistics journal January 1952
An Analysis of Variance Test for Normality (Complete Samples) journal December 1965
Experimental Statistics. journal June 1964
Experimental Statistics journal November 1965

Figures / Tables (15)

Table 1(p. 38)table Table 1
Table 2(p. 39)table Table 2
Figure 1(p. 40)figure Figure 1
Figure 2(p. 41)figure Figure 2
Figure 3(p. 42)figure Figure 3
Figure 4(p. 43)figure Figure 4
Figure 5(p. 44)figure Figure 5
Figure 6(p. 45)figure Figure 6
Figure 7(p. 46)figure Figure 7
Figure 8(p. 47)figure Figure 8
Figure 9(p. 48)figure Figure 9
Figure 10(p. 49)figure Figure 10
Figure 11(p. 50)figure Figure 11
Figure 12(p. 51)figure Figure 12
Figure 13(p. 52)figure Figure 13

Similar Records

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
Advanced Test Reactor
Verification and Validation
Uncertainty Quantification
95% Confidence Rule
Reliability Factor