INTERNATIONAL CONFERENCE ON NUCLEAR DATAFOR SCIENCE AND TECHNOLOGYMay 19 - 24, 2019 • Beijing, China

CONFERENCE PROGRAM& ABSTRACT BOOK

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Tue May 21

14:00-17:50

Room 305

Topic Track: Nuclear Data Application
Session Title: Application in Nuclear Reactor 3

Chair: Tamara Korbut
14:00

I158

14:30

R159

14:50

R160

15:10

R161

15:30

R162

A New Reference Database for Beta-delayed Neutron Data for Applications/ Paraskevi Dimitriou (International Atomic Energy Agency, On behalf of theIAEA CRP, Austria)
On-the-ﬂy Temperature-dependent Cross Section Treatment Under Extreme
Conditions in RMC Code / Lei Zheng (Department of Engineering Physics,Tsinghua Uni., China)Evolution of the Importance of Neutron-induced Reactions Along the Cycleof An LFR / Pablo Romojaro (CIEMAT, Spain)
Benchmarking the New ENDF/B-VIII.0 Nuclear Data Library for OECD/NEA
Medium 1000 MWth Sodium-cooled Fast Reactor / Donny Hartanto (Uni. ofSharjah, UAE, United Arab Emirates)
Measurement of Temperature-dependent Thermal Neutron Spectrum in
CaH2 Moderator Material for Space Reactor Using TOF Method / Jaehong
Lee (Ins. for Intergrated Radiation and Nuclear Science, Kyoto Unversity, Japan)

Break

15:50

Topic Track: Nuclear Data Application
Session Title: Application in Nuclear Reactor 4

Chair: Ping Liu
16:10

R163

16:30

R164

16:50

R165

17:10

R166

17:30

R167

Production and Veriﬁcation of the Compressed Depletion Data Library for
Neutronic Analysis / Yunfei Zhang (Harbin Engineering Uni., China)
Benchmarking the New ENDF/B-VIII.0 Nuclear Data Library for the FirstCore of Indonesian Multipurpose Research Reactor (RSG-GAS) / Donny Hartanto (Uni. of Sharjah, UAE, United Arab Emirates)
Nuclear Data Sensitivity and Uncertainty Analyses on the First Core Criticalityof the RSG Gas Multipurpose Research Reactor / Peng Hong Liem (Tokyo CityUni., Japan)
On the Impact of Nuclear Data Uncertainties on LWR Neutron Dosimetry
Assessments / Dimitri Rochman (Paul Scherrer Institut, Switzerland) / Speaker:
Erwin Alhassan
Nuclear Data Sensitivity Analysis and Uncertainty Propagation in the KYADJ
Whole-core Transport Code / Qu Wu (Nuclear Power Ins. of China)

15

R161 Benchmarking the New ENDF/B-VIII.0 Nuclear Data Libraryfor OECD/NEA Medium 1000 MWth Sodium-cooled Fast
ReactorDonny Hartanto1 , Bassam Khuwaileh1 , Peng Hong Liem2,3
1. University of Sharjah, UAE
2. Tokyo City University (TCU), Japan3. Nippon Advanced Information Service (NAIS Co., Inc.), Japan
This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for theOECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic
fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo codes Serpent 2
will be used as the calculation tools. Several nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0
will be included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k-eﬀ,beta-eﬀ, sodium void reactivity (ρNa), Doppler constant (ρDoppler), control rod worth (ρCR), andradial power distribution.

of Temperature-dependent Thermal Neutron
R162 Measurement
Spectrum in CaH2 Moderator Material for Space Reactor
Using TOF Method
Jaehong Lee1 , Tadafumi Sano 1 , Jun-ichi Hori 1 , Rei Kimura2 , Takayuki Sako 2 , Akira Yamada2 ,Jun Nishiyama3
1. Institute for Intergrated Radiation and Nuclear Science, Kyoto Unversity
2. Toshiba Energy Systems & Solutions
3. Laboratory for Advanced Nuclear Energy, Institute of Innovative Research, Tokyo Institute ofTechnologyIn order to power the space exploration such as human missions to the moon and Mars, the small hightemperature nuclear reactor using a solid moderator has been proposed as a space power system. As asolid moderator, the calcium-hydride (CaH2) having a high melting point, 816 ◦ C, is suggested for thehigh-temperature operation.
Because neutronics characteristics is changed by the increase of temperature, the reactivity of the CaH2
moderated space reactor is greatly contributed to the temperature-dependent thermal neutron spectrum
in the moderator. Nevertheless, the evaluated nuclear data and experimental data of the CaH2 is notsatisﬁed in quality and quantity. Therefore, in order to accurately design the reactor, it is necessary toexperimentally conﬁrm the temperature-dependent thermal neutron spectrum in the CaH2.
In the present study, we have focused on the measurement of the temperature-dependent thermal neutron
spectrum in the CaH2. To obtain the temperature-dependent thermal neutron spectrum, we have carriedout the neutron scattering experiment using the time-of-ﬂight (TOF) method at the Kyoto University
Institute for Integrated Radiation and Nuclear Science - Linear Accelerator (KURNS-LINAC). In presentexperiment, we raised the temperature of the CaH2 from a room temperature to 500 ◦ C, and obtained
the change of the thermal neutron spectrum for the increase of temperature.
The obtained temperature-dependent thermal neutron spectrum in the CaH2 is compared with the calculated result using the Monte-Carlo simulation with JENDL-4.0.
99

and Veriﬁcation of the Compressed Depletion
R163 Production
Data Library for Neutronic Analysis
Yunfei Zhang, Qian Zhang, Qiang ZhaoHarbin Engineering University
The depletion calculation based on the explicit depletion library can accurately describe the revolutionof reactor composition, However, it requests large computing cost in reactor physics calculation. In
this paper, a high-ﬁdelity compression depletion library is produced based on quantitative signiﬁcanceanalysis. In PWR problem, the eﬀects of each unit compression operation on neutron absorption, neutron
output, and target nuclide nuclear number density are investigated. For the compressed library, a series ofbenchmark problems are used to veriﬁcation. The results show that the simpliﬁed library can signiﬁcantly
reduce the burn-up chain scale and save computing resources, while meeting the calculation accuracy
requirement.

R164 Benchmarking the New ENDF/B-VIII.0 Nuclear Data Li-brary for the First Core of Indonesian Multipurpose Research
Reactor (RSG-GAS)

Donny Hartanto1 , Victor Gillette1 , Tagor M. Sembiring2 , Peng Hong Liem3 , 4
1. University of Sharjah, UAE
2. National Nuclear Energy Agency (BATAN), Indonesia
3. Tokyo City University (TCU), Japan
4. Nippon Advanced Information Service (NAIS Co., Inc.), JapanThe Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS)
is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using lightwater as coolant and moderator, and beryllium as reﬂector. The benchmark of the 1st criticality coreof RSG GAS using diﬀerent nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, andJEFF-3.1 have been performed in the previous work and compared with the experiment result. In thiswork, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will
be used and the important neutronics parameters such as multiplication factor, kinetics parameters, andﬁssion reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against theprevious work and the experiment result.
100Data Sensitivity and Uncertainty Analyses on theR165 Nuclear
First Core Criticality of the RSG Gas Multipurpose ResearchReactor
Peng Hong Liem1,2 , Zuhair Zuhair3 , Donny Hartanto4
1. Tokyo City University
2. Nippon Advanced Information Service (NAIS Co., Inc.), Japan
3. National Nuclear Energy Agency (BATAN), Indonesia
4. University of Sharjah, UAE
The results of criticality, sensitivity and uncertainty (S/U) analyses on the ﬁrst core criticality of theIndonesian 30 MWth Multipurpose Reactor RSG GAS (MPR-30) using the recent nuclear data libraries(ENDF/B-VII.1 and JENDL-4.0) and analytical tools available at present (WHISPER-1.1) are presented.
Two groups of criticality benchmark cases were carefully selected from the experiments conducted during
the ﬁrst criticality approach and control rod calibrations. The C/E values of eﬀective neutron multiplication factor (k) for the worst case was found around 1.005. Large negative sensitivities were found in (n,γ)
reaction of H-1, U-235, Al-27, U-238 and Be-9 while large positive sensitivities were found in U-235 (totalnu and ﬁssion), H-1 (elastic), Be-9 (free gas, elastic) and H-1 S(α,β) (lwtr.20t, inelastic). The energy
dependency of the sensitivities were also presented and discussed in detail. The S/U analysis results
concluded that the uncertainties of k originated from the nuclear data were found around 0.6% which
covered well the [C/E-1] values. Diﬀerences in the sensitivities amongst the two nuclear data libraries
were also identiﬁed, and recommendation for improving the nuclear data library was given.

the Impact of Nuclear Data Uncertainties on LWR NeuR166 On
tron Dosimetry Assessments
Dimitri Rochman, Alexander Vasiliev, Hakim Ferroukhi
Paul Scherrer Institut
In this work, an overview on the relevance of the nuclear data (ND) uncertainties with respect to the Light
Water Reactors (LWR) neutron dosimetry is presented. The paper summarizes ﬁndings and observations
accumulated on the basis of several studies realized at the LRT (former LRS) laboratory of the Paul
Scherrer Istitute over the past decade. The studies were done using the base LRT calculation methodology
for dosimetry assessments, which involves the neutron source distribution representation, obtained basedon validated CASMO/SIMULATE calculation models, and the consequent neutron transport simulations
with the MCNP code. The methodology was validated using as the reference data the results of numerous
measurement programs realized at Swiss NPPs and at PSI Hotlab. Namely, the following experimental
programs are considered in the given overview: PWR RPV scraping tests and PWR ”gradient probes”
post irradiation examination (PIE) data and BWR reactor fast neutron ﬂuence monitors dosimetry PIE
data.
The assessment of the nuclear data related uncertainties was performed with the help of calculation
tools and methodologies developed and maintained at LRT, which in principle can cover the complete
calculation chain from the cycle-speciﬁc core-follow calculations to the neutron transport and ﬁnally tothe dosimetry reaction rates calculations. However, so far the above uncertainty components were mainly
analyzed separately from each other in order to facilitate the separate tools development and associatedV&V studies. When appropriate, the results of the cross-veriﬁcation of the diﬀerent ND uncertainty
quantiﬁcation tools and techniques realized at LRT will be discussed too.
Furthermore, the ﬁndings on which particular neutron induced reactions contribute dominantly to theoverall ND-related uncertainties will be presented. In addition, the magnitude of the ND-related uncertainties will be assessed against other types of typical modeling and experimental uncertainties basedon the information available at LRT. Thus, the collected results should be suﬃciently representative toassess the overall impact of the various nuclear data uncertainties with respect to the practical reactor
dosimetry applications and provide relevant feedback to the nuclear data evaluators.
101