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Report on the progress of CRP 15009 research contract
Hongchun Wu, Liangzhi Cao*
[email protected]
Department of Nuclear Engineering, Xian Jiaotong University,
China
1. Background information (areas of experience and expertise in
solution reactor
design, analysis, construction, operation, maintenance, waste
management, etc. as
well as similar information related to isotope extraction from
solution reactor
operation).
NECP (Nuclear Engineering Computational Physics) lab of Xian
Jiaotong
University has been focusing on neutron transport calculation
method and its
application in nuclear engineering for several decades.
On the one hand, many neutron transport solvers have been
developed, such as:
Auto-MOC : based on the Method of Characteristics and the
customizationof the AUTOCAD software;
TEPFEM : based on the Finite Element Method and the Pn method;
LESFES: based on the Finite Element Method and the Sn method; DNTR:
based on the triangular nodal method and the Sn method; TEMTD:
based on the transmission probability method; DDPM : based on the
discrete direction probability method; and so on.On the other hand,
many applications of these solvers have been implemented.
Aqueous homogenous solution reactor, which is regarded as a
promising candidate forproducing medical isotopes, is one of the
objects of these applications.
2. Detailed information on current and planned research
activities including specific,
expected outputs.
A software package for the core design, fuel management
calculation and safety
analysis of the aqueous homogenous solution reactor has been
developed in NECP lab.
This software package contains four parts:
few-group constants computation code PANTHER three-dimensional
in-core fuel management calculation code FMSR three-dimensional
core neutronic kinetics analysis code DNTR-T and point-reactor
kinetics analysis code POINT-R.
PANTHER begins with the 69-group WIMS library released by the
IAEA and
performs the transport calculation for the complex geometry of
the solution reactor. It
supplies few-group constants for 3D steady state calculation
code FMSR and 3D
transient state computation code DNTR-T. FMSR initializes
transient analysis
calculation for DNTR-T and POINT-R except steady state fuel
management
calculation. The detailed theoretical model of the PANTHER/FMSR
can be found in
our published paper (FMSR: A code system for in-core fuel
management calculation
of aqueous homogeneous solution reactor, Nuclear Engineering and
Design, 2010).
DNTR-T performs efficient 3D transport kinetics calculation by
combines improved
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quasi-static method (IQS) and triangular nodal Sn method.
POINT-R employs
analytical method to solve the point reactor kinetics equation
by considering the void
and temperature feedback. All these four code systems have been
verified against
some benchmark problems. Further validation is undertaken.
As the future research plan, low enrichment uranium (LEU) fueled
aqueoushomogenous reactor (AHR) core neutronic design will be
carried out by using the
code system developed above. Firstly, the design objectives of
the LEU fueled AHR
must be specified by investigation or communication with IAEA.
For example, the
power level, the power peak factor, the reactivity feedback
coefficients and generating
rate of the medical isotopes et al, must satisfy some criteria.
Secondly, an reference
core design will be obtained by following the design criteria.
Thirdly, the reference
design will be improved and optimized.
3. Ideas related to opportunities for collaboration (per the
attached draft agenda
beginning at 16:00 on 24-February)In our opinion, the
verification and validation of the codes are of critical
importance and bring a lot challenges to the code system
development. Currently, the
code system was only verified against some simple benchmark
problems, but not
experimental data. It is difficult to confirm that the codes are
exactly accurate with
practice and absolutely reliable. In order to specify the
difference between numerical
simulation and actual operation, many experiments will be needed
or some other ways
to verify the codes need to be designed. So, we propose the
collaboration on the
development of solution reactor benchmarks under the IAEA
support. All participants
can join the program to give their own solutions to a series of
common benchmark
problems.
In spite of that, some basic works will be needed. For instance,
the 69-group
WIMS library does not contain some important medical isotopes
such as Mo-99. The
present point burnup model can not lead us to the fine
distribution of the all the
isotopes in reactor core.
4. Your current work plan (activities and schedule).
2010.2-2010.5 Further verification and validation of the codes
2010.6-2010.7 Preliminary core design of the LEU fueled AHR
2010.8-2010.9 Neutronics analysis and optimization of the LEU
fueled AHR 2010.10 Project summary
