PSI Scientific Report 2006
Research focus and highlights – Nuclear energy and safety
computed FNF values, obtained using the currently developed 500 0
CASMO-4/SIMULATE-3/MCNPX calculation scheme, are compared against the estimates based on activity measurements
-500
of the scraping samples in Figure 3. The present calculational
-1000
results lie within ± 5 % of the reference values, i.e. well with
-1500
the declared experimental uncertainty of ± 10 %, and can be
FR A U NC K E /B N FL FI N LA U K JA / S ND PA JA ER P N JA /JA AN CO PA ER / J N N /J I/JE ES AE N D R I/J L3 EN . 3 D L G ER 3.2 M U SA AN /H Y U El SA US lO A /T S / U SA RIT SAS / T ON 2h R IT KE O N O SW N IT NE ZE W T R LA N D
Δ kinf, pcm
1000
considered to be very satisfactory.
Conclusions
Figure 2: Evaluated results for BUC-IID benchmark.
As a result of the considerable advance in computing capabilities, and the quality of neutron data (but also as a result
Studies related to burn-up credit
of increased safety demands), there is a global trend to intensify the usage of MC codes in neutron transport calculations.
Taking credit for fuel burn-up of discharged fuel assemblies
The development of current R&D activities at PSI exemplifies
enables significant cost optimizations for the storage and
this trend. In particular, appropriate application of the MCNPX
transportation of spent fuel to be made. Ongoing studies
code within the STARS project has made it possible to achieve
aimed at assessing MC-based computational accuracies in
a significant enhancement of accuracy for several different
this context, necessary for BUC implementation into CSE
types of NPP-relevant analyses, as illustrated by the examples
methodology. As an example, results of participation in the
given here.
OECD/NEA BUC calculational benchmark for PWR-UO2 fuel assembly burn-up and criticality calculations [3] are presented in Figure 2. The value shown, for each participant, is finite neutron multiplication factors kinf from the mean values calculated from the submissions by all participants. The PSI (Switzerland) results, obtained with the MONTEBURNS2.0/ MCNPX2.4.0/ORIGIN2.1 code system in conjunction with the related data libraries, correspond very closely to the overall mean value.
Fast neutron fluence analysis
4.0
Neutron fluence, 1018 n/(cm2)
the average deviation (for the 23 cases calculated) of the in-
3.8
Scraping test - 93mNb, ±σ exp C4/ S3/ MCNPX/ JEF-2.2, ±σ MC
3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 90
100 110 120 130 140 150 160 170 180
θ, °
Radiation-induced damage in metals, which may be approximated as a function of the FNF, is an important safety-related factor influencing plant lifetime. A novel methodology for
Figure 3: FNF azimuthal distribution over RPV inner surface at core mid-height after 10 reactor cycles.
accurate modelling of the FNF accumulated in a LWR RPV has been developed and validated in support of the ageing as-
References
sessment of Swiss NPPs [2]. The methodology is based on the
[1] E. Kolbe, A. Vasiliev, M.A. Zimmermann. “Assessment of
transfer of deterministic CASMO-4/SIMULATE-3 core-follow
Standard Point-wise Neutron Data Libraries for Criticali-
results (power distribution, fuel compositions), representing
ty Safety Analysis with a Monte Carlo Code”, In Proc. of
the actual reactor operational history, into a three-dimen-
ANS Topical Meeting on Reactor Physics “Advances in
sional volumetric (pin-by-pin) fixed neutron source for ex-core
Nuclear Analysis and Simulation”, “Physor-2006”, Van-
neutron transport simulations using MCNPX.
couver, BC, Canada, September 10-14, 2006.
Appropriate qualification of a methodology for FNF calcula-
[2] A. Vasiliev, H. Ferroukhi, M.A. Zimmermann. “CASMO-4/
tions should include comparisons with measurements from
SIMULATE-3/MCNPX Analysis of a Reactor Pressure Ves-
operating reactors. To this purpose, ‘scraping test’ data from the RPV of a Swiss NPP have been used as a source of validation for the entire sequence of steps in the FNF analysis. The
sel Scraping Test”, ibid. [3] A. Barreau, “Burn-up Credit Criticality Benchmark Phase II-D”. ISBN 92-64-02316-X, NEA No. 6227, OECD 2006.
83