COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES

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					2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY
Beijing, CHINA, September 22-24, 2004                                                               #Paper C22




COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES

       FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED HTR
                                                REACTORS

                               T. RADEMER, W. BERNNAT and G. LOHNERT
    Institute of Nuclear Technology and Energy Systems (IKE), University of Stuttgart, Pfaffenwaldring 31, 70569
                                                Stuttgart, Germany


ABSTRACT: The Pebble Bed Reactor plant is a concept of an inherently safe nuclear power plant.
This paper presents the coupling of the SIEMENS INTERATOM-developed and licensed 2D neutron
diffusion code KIND with the commonly used thermal-hydraulics code THERMIX to simulate very
fast transients of High Temperature Reactors (HTR). This is possibly by using more up-to-date
thermal-hydraulic models and a more detailed heterogeneous particle model for the fuel kernels.
Furthermore, THERMIX is implemented in the steady-state ZIRKUS modular code system which
provides necessary initial conditions for the transient code and nuclear data base (macroscopic cross
sections) as well.

Additionally, a Graphical User Interface (GUI) called SIMPLAT is designed to simplify the input
creation and the results analyses for ZIRKUS as well as for KIND. Moreover, the 1D-KIND transient
code is coupled with the graphic development software LabView1 to prove the possibility of a runtime
interactive intervention of the user during the calculation. This interactive intervention enables to
introduce parameter changes while the code is already running.
To study the influences of reactivity transients which causes effects that are as fast as the relaxation
time for heat exchange between the fuel particles (~ 10 ms) and the surrounding graphite. Results of
heterogeneous and homogeneous simulations are compared and show an erroneous overestimation of
the maximum power (by one order of magnitude) for the homogeneous model, since the Doppler
effect is calculated based on the homogenized and not the real fuel temperature.


0. INTRODUCTION
The correct simulation of power distribution, flow and heat transport is of great importance for the
next generation high temperature gas-cooled reactors (HTR) such as the “Pebble Bed Modular Reactor
(PBMR) (South Africa)” [1] or the HTR-MODUL reactor developed by SIEMENS/Interatom
(Germany) [2]. Furthermore it is very important to study the reactor behavior in the case of an accident
during the design phase to avoid a failure in design. Future reactors should be easier to build, simpler
in their maintenance, inherent safe, and more economical as described by Zanocco in [3]. To achieve
this objectives common tools have to be improved to perform the safety related analysis for various
designs, e.g. reactivity transients or LOCAs.
At the Institute of Nuclear Technology and Energy Systems (IKE), Germany, we are using the kinetic
and dynamic neutronic code called RZKIND which is using an own thermal-hydraulic routine called
THERM to calculate the temperature fields. This 2-dimensional code was originally developed to
simulate the transient behavior of the German pebble bed HTR-MODUL reactor for different
accidental scenarios and was licensed by the German licensing authorities (TÜV). This code is also
available in a 1D version (ZKIND,WKIND). The code contains a detailed heterogeneous model for

1
    Copyrights for “LabView “ by National Instruments



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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22

the heat transport from coated particle fuel to moderator which can simulate very fast reactivity
transients much better than a homogeneous model which will be described later on.
For the preparation of cross-sections in multigroup approximation the modular program system
ZIRKUS developed by Framatome ANP GmbH (former SIEMENS-KWU) was used. It was designed
for the realization of various reactor physics calculations for the HTR cores with pebble fuel [4].
It can be used to prepare a cross-section data base for transient calculations by representing the few
group cross-sections as a function of different parameters like fuel temperature, moderator temperature,
reflector temperature, Xe-density, H2O-density (if water ingress is expected) and control rod positions,
which is needed by the transient KIND codes.
Additionally, a Graphical User Interface (GUI) called SIMPLAT is in the design phase by our Dept. of
Knowledge Engineering and Numerics to simplify the input creation and the results analyses for
ZIRKUS as well as for RZKIND. Originally the ZIKRUS input files which are necessary for a
calculation had to be prepared by hand. SIMPLAT makes it easier to start and manage the manifold
runs of ZIRKUS.
Moreover the 1D-ZKIND of the transient KIND code family is coupled with a program development
environment called LabView to prove the possibility of a runtime interactive intervention of the user
during the calculation. This interactive intervention enables to introduce parameter changes while the
code is already running.


1. DESCRIPTION AND CURRENT DESIGN
The aim of this work is the development of a simulation package to integrate safety analysis through
accident evaluation during early basic engineering stages of the reactor design. One step on this way is
the coupling of the thermal-hydraulic code THERMIX/KONVEK with the neutronics code RZKIND
and the use of a more detailed heterogeneous particle model for the fuel kernels. The importance for
the simulation of very fast transients will be shown. A further step is the implementation of a runtime
interactive intervention graphical user interface.
1. 1. Coupling RZKIND with THERMIX/KONVEK
The RZKIND code work on the basis of macroscopic one group cross-sections, which are
implemented in an associated library as functions of feedback parameters in a polynomial
representation. This library is a result of manifold ZIRKUS calculations. Fig. 1 displays the data
transformation from ZIRKUS to RZKIND. Many modules have to be used to prepare data for transient
calculation with the KIND codes. Like ZIRKUS these modules have data exchange via a general data
base by the mass storage archiving and retrieval system MARS [4;6](see Fig.1). RZKIND treats a 2D
cylindrical R-Z geometrical structure as well as THERMIX.
The KIND code itself consists of three main parts, which are executed sequentially within each time
step which is displayed in Fig. 2.
Firstly, the neutronic part, to calculate the neutron flux density and the power density in the reactor
core using a one group neutron diffusion equation, equations for the delayed neutrons in the usual
representation with six groups, equations for the fission product poisoning through Xe-135 (two
isotopes) and Sm-149 (chain with six isotopes). Furthermore special treatment for the decay heat (so
called pseudo emitters) is implemented.
The second part is the thermal-hydraulic part to calculate the temperatures of the fuel elements (solid
material) and the coolant gas (helium) in the pebble bed core and in the case of the 2D KIND in the
reflector regions. The determination of the helium temperatures in the pebble bed assumes isolated
channels with forced streaming.




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HTR2004                                                                              Beijing, CHINA, 2004. 9.



                                      ZIRKUS
                                 equilibrium calculations

                                  ZIRKUS modular
                                      sequences for
                                       variations of
                                        parameters                  M
                                                                    A
                                      KONTR                         R
                                  1 group cross sections
                                      and boundary                  S
                                   conditions (leakage)
                                  for varied parameters
                                                                    D
                                         VPF                        A
                                       polynomial
                                       expansions                   T
                                    of cross sections               A
                                      and leakage

                                  KINZL, YIKO                       B
                                     DEVA                           A
                                    yields and kinetic
                                       parameters                   S
                                                                    E
                                     DATGEN
                                    compile RZKIND
                                       data file


                                     RZKIND
                                        data file




                     Figure 1: Flow chart for preparation of RZKIND data base

The third part characterizes the neutronic/thermal hydraulic feedback which evaluates modified cross-
sections for the neutronic part due to changed parameters. The main parameters are the fuel
temperature and the moderator temperatures in the pebble bed, the distribution of the xenon
concentration in the fuel elements, the distribution of the water steam concentration in the coolant
(accidental water ingress) as well as the position of the control rods and the shutdown elements [4].
The cooling gas passing the pebble bed is treated as flowing in independent radial channels with equal
mass flow densities. Only convective heat transfer at the side reflector is considered. This calculation
model permits flow rates down to 10% of the nominal mass flow. This restrictions makes it necessary
to implement an other thermal-hydraulic tool. Therefore, THERMIX/KONVEK is coupled in a first
version with the 2D-RZKIND code, to a coupled version called READY.
THERMIX/KONVEK simulates the fluid flow alone with the temperature distribution in the fluid and
conducting solids based on a porous media approach which involves the solution of the time
dependent energy conservation equation for solids and quasi-steady mass and energy conservation
equation together with a simplified momentum equation (Ergun-type with buoyancy) for the gas. For
further information see [5] by Nader Ben Said.




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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22




    Figure 2: Principle flow chart of the coupled version of THERMIX and KIND dash line the new
    THERMIX calculation and black line the original THERM path. This coupled version is called
                                                READY.

In addition there is an appropriate set of empirical correlations for the friction and the pressure drop,
modeled by KTA-rule2 3102.3. The properties of the fluid are implemented in accordance to KTA-
rule 3102.1. The pressure drop caused by the pebble bed is modeled in accordance to KTA-rule 3102.2
[13; 14; 15].
For all these reasons the THERMIX/KONVEK module is preferred to obtain more accurate solutions;
e.g. if natural convection plays a role in a LOFC accident, with flow rates below the limit of 10 % of
the nominal mass flow which restricts the THERM routine.
Fig. 2 shows the principle sketch of the coupling between the transient code KIND and the thermal-
hydraulic code THERMIX/KONVEK. The sketch shows that the former stand-alone code THERMIX
is now controlled by the KIND system, but do not replace the original THERM routine; it serves as an
alternative, e.g. if natural convection is important in an accidental scenario.
Due to the fact that KIND as well as THERMIX have an own time-step control it had been necessary
to give one code the leadership. In this case it is KIND which gives the time-step width to
THERMIX/KONVEK.


2
 KTA rules are Safety Standard from the German Nuclear Safety Standards Commission (KTA)
In the internet: http://www.kta-gs.de/



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HTR2004                                                                                                                         Beijing, CHINA, 2004. 9.


                   1600

                                        T-M-0s
                   1400
                                        KIND-T-M-0s


                   1200


                   1000
   ]
   Temperature [




                    800


                    600


                    400


                    200


                      0
                          0   50         100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1000
                                                                                 Height [cm]


 Figure 3: Temperature distribution over the heights of the HTR reactor model at the beginning of the
transient calculation. The black boxes represent temperature distribution of the KIND calculation with
    the original THERM routine and the red rhombi the results of the calculation with THERMIX

                                                        80



                                                        70
                                                                             R Z K IN D (b lu e )
                                                                             c a lc u la tio n

                                                        60



                                                        50
                                   Relative Power [-]




                                                        40



                                                        30



                                                        20                             P a rtic le M o d e ll o f
                                                                                       Z K IN D re d


                                                        10



                                                         0
                                                             0   1   2   3         4          5         6           7   8   9         10
                                                                                         T im e [s ]


        Figure 4: Relative power for the use of the homogeneous model and the heterogeneous
                   particle model. Calculated with the kinetic codes RZKIND and ZKIND




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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22

Furthermore, the power-density calculated by KIND has to be moved to THERMIX to calculate the
temperature field which is used by the KIND code for the next time-step. Fig. 2 shows that the
calculation process is divided in two main parts. The stationary branch and the transient branch.
Fig. 3 turn out a comparison of the moderator temperature calculated with the THERM module in
RZKIND and a calculation with THERMIX, for identical initial conditions after the stationary
calculation just at the beginning of the transient calculation. It points out only few differences. The
differences in the range of these curves is a result of different net grids used in RZKIND and
THERMIX and a different consideration of the extensions in the model. For the future both codes
should use the same basic grid net. Further information on recriticality calculations of HTRs can be
found in [7; 12].
1. 2. Coupling of WKIND with TRANSIMOLA
To study the possibility of coupling the neutronic code WKIND with a LabView [8] based graphical
user interface (GUI), TRANSIMOLA was developed. Principle sketch of the data flow can be seen in
Fig. 5.
TRANSIMOLA is a user environment to control input data and also the output results.
The reasons are to prove if simulations can be performed faster and more effective with an interactive
intervention of the user while the calculation is running. The user has faster feedback with the results
compared with the normal calculation process. In the normal calculation process starting conditions
and the transient behavior like control rod withdrawal or changes of the time-dependent helium inlet
temperature as pointwise linear function have to be defined in an input file before the calculation can
be started.
Then the cross-section data base produced by manifold ZIRKUS calculations is linked to the code and
the transient calculation runs. After a problem depending time period the results are written in an
ASCII type output file ready to be analyzed. This process requires much time from the definition of a
transient problem to the interpretation of the results.




                Figure 5: left: Principle sketch of the normal calculation process
                right: Principle sketch of the interactive intervention during simulation

In the case of using TRANSIMOLA the starting conditions have to be specified as in the customary
calculation process, but the new graphical user interface makes it possible to notice wrong inputs
much faster and save time, because the output is directly linked to TRANSIMOLA and gives a real-
time feedback to the user. Mistakes can be recognized and calculation runs can be stopped. In addition,
TRANSIMOLA allows to change parameters like the control rod position or the helium inlet mass
flow rate during the calculation.
This allows engineers to change parameters in such a way that the simulation system reacts like a
control desk in a reactor and can be used as an educational tool for operators.




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HTR2004                                                                            Beijing, CHINA, 2004. 9.


For these reasons the system is developed with the requirement to run on spatial distributed computers
(see Fig. 6). It is possible to start the neutronic code WKIND3, e.g. on a UNIX system and the
Graphical user interface on a separate Windows 2000 PC.
The data is transferred back and forth over a Data Wheel (Fig. 6), which can be interpreted as an hard
disk space which can be reached via a UNIX system and the local PC system of the IKE. Fig. 7
present the user environment of the TRANSIMOLA graphical user interface (GUI), where the input
and output variables can be chosen and the transient defined, like for instance the helium inlet
temperature, the helium mass flow rate or the control rod position. Fig. 8 and 9 shows sample output
screens on TRANSIMOLA.




    Figure 6: Principle sketch of data flow between 1D KIND and Transimola the LabView user
     interface. The data is exchanged via the Data Disk an storage file system at the Computer
                                          Center Stuttgart.




Figure 7: The TRANSIMOLA graphical user interface (GUI) made with the aid of LabView to control
  the WKIND calculation. You can see the input parameters and the output parameters the user can
                                              chose.


3
 Beside the UNIX version there exists also a WINDOWS version for PC, WKIND is a 1D Version of
KIND


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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22

1. 3. Examples
Beside the breakdown of the primary cooling circuit the withdrawal of the control rods is an accident
scenario which has to be regarded. In this example a hypothetical extremely improbable reactivity
accident is simulated for the HTR-MODUL reactor developed by SIEMENS/Interatom.
Fig. 8 and Fig. 9 show the accidental situation in which all the control rods withdraw with a velocity
of 1 m/s out of the core up to the end position. There are two SCRAM values implemented simulating
the safety systems of the reactor. One checks the relative power/relative flux and the other the helium
outlet temperature. If the relative power reaches 1.2 times the nominal value or the helium outlet
temperature exceeds a positive temperature increase of 50 Kelvin than a SCRAM signal is given. The
nominal thermal power is 200 MW.
In our case the SCRAM signal is activated; however , the control rods are not stopped, but get stuck at
the end position and will not be inserted - as foreseen. Only the helium mass flow is reduced by
controlling the blower (Fig. 8, below).
The nominal starting position of the control rods is 240 cm inserted in the active core-zone. The
control rods withdrawal with a velocity of 1 m/s out of the core up to the end position by –100 cm.
This happens in 3.4 seconds (see Fig. 8 above). The withdrawal of the control rods causes an increase
of the reactivity of 1.2 % and effects an increase of the relative power demonstrated in the upper
section of Fig. 9. The first scram activates promptly and causes a contrary effect through the
decreasing helium mass flow shown in Fig. 8. The result is a peak curve of the relative power.




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HTR2004                                                                               Beijing, CHINA, 2004. 9.




                                              Spos [cm]: control rod position in cm




                                              r.m: relative He mass flow rate




Figure 8: Fast control rod withdrawal by constant helium inlet temperature and constant helium mass
   flow rate up to the SCRAM. r.m. is the relative mass flow of helium and Spos is the Control rod
                                           position in cm.




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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22




                                                 r.P: relative thermal total power




                                                 Taus [K]: He outlet temperature
                                                 T,B [K]: fuel temperature




Figure 9: In the upper picture the relative thermal power (r.P.) is shown as a result of the fast control
rod withdrawal with constant helium inlet temperature and constant He mass flow up to the SCRAM.
   In the lower picture the helium outlet temperature and the fuel temperature measured in Kelvin.




                                                -- 10 --
HTR2004                                                                               Beijing, CHINA, 2004. 9.


1. 4. SIMPLAT
SIMPLAT is a framework under development by our Dept. of Knowledge Engineering and Numerics
to utilize the integration of simulation codes like KIND and ZIRKUS in simulation applications (see
Fig. 10). An application consists out of a number of services to allow the user to run a simulation
through a certain context. It generates the data sets and control the input generation for all the program
modules which are necessary to run a complete transient simulation, archives the outputs and presents
the results.
By the help of SIMPLAT the work with complex simulations is easier and more error-free and
therefore cheaper. Figure 11 shows a typical data sheet of SIMPLAT.




Figure 10: Data Flow Chart of the SIMPLAT SYSTEM combined with the ZIRKUS and KIND System




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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22




Figure 11: Typical data sheet of the Info Session of SIMPLAT which contains often changed physical
                                             parameters

The motivation for this project results in on the claims out of the engineering practice that normally
simulation tools could only be handled by specialist. SIMPLAT offers an option to make the
knowledge of this experts transparent and for users more understandable. Simulations tools hence
become available as a service at the working place.
Examples for the outputs in a first version are given in Figure 10 a-d to show that the system is
principally working.




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HTR2004                                                                             Beijing, CHINA, 2004. 9.




(a)                                                 (b)




(c)                                                 (d)


Figure 12:      Examples of different results which are produced by the help of the SIMPLAT system.
                 (a) shows an example form the neutron flux density of energy group 1
             (b) the solid temperature in °C over the radius of the cylindrical HTR reactor
                                 (c) the decay heat vs time im MW/m³
                              (d) the power density in the core in MW/m³


2. CONCLUSIONS AND PROSPECTS
The aim of this work is the development of a simulation package to integrate safety analysis through
accident evaluation during early basic engineering stages of the reactor design. Therefore the approach
of coupling available codes has been to compile a system to simulate neutronic and thermal-hydraulic
behavior of the core in the future the complete primary circuit of an HTR inclusive the core itself is
envisaged. Stand-alone programs for the transient reactivity calculations and thermal-hydraulics are
existing. Research on the capabilities of the codes shows the necessity of improvements. One step has
been the coupling of the thermal-hydraulic THERMIX/KONVEK with RZKIND to exempt the
neuronic code KIND from the limitations of the THERM routine. An other step has been to use a
more accurate heterogeneous model for heat transfer of then fuel kernels to the surrounding graphite
for very fast transients, because the homogeneous model overestimates the maximum power by one



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COUPLING OF NEUTRONICS AND THERMAL-HYDRAULICS CODES FOR THE SIMULATION OF TRANSIENTS OF PEBBLE BED
HTR REACTORS                                                                                   #C22

order of magnitude, since the Doppler effect is calculated based on the homogenized and not the real
fuel temperature.
With reference to the paper of A. Walter [10] which couples the CFD code FLOWNEX4 with the
KIND the first step to a complete system is done.
To simplify the use of this codes the two systems SIMPLAT and TRANSIMOLA are under
development. TRANSIMOLA as a graphical runtime interactive intervention tool and SIMPLAT as
an development environment and management system for ZIRKUS-KIND calculations.
TRANSIMOLA shows in combination with the 1D WKIND code that it can accelerate to obtain
solutions through its interactivity by minimizing defective runs and easy output analyses. Changing of
cryptic input files which is an error-prone work, can be reduced. Furthermore, the user can intervene
during a calculation by changing temporal behavior of the reactor in the simulation, e.g. by changing
the control rod velocity or He-mass flow rates. This can be done at any time during the calculation.
This ability can be used to simulate simple circuits connected to the core.
The neutronic codes are verified by re-calculation of experiments and AVR operating data [12] as well
as by comparison with other calculations and reference solutions [7]. Most of the models used in these
programs are proven for usual design calculations. In the actual version SIMPLAT manages ZIRKUS
and KIND runs; it is still under development. For the future more improvements will be envisaged, e.g.
to transfer the 2D KIND into a 3D version and using multi-group data instead of 1-group data used so
far. Also a coupling with the CFD code CFX is planned.




4
 A thermal-fluid CFD code from M-Tech Industrial (Pty) Ltd., PO Box 19855 Noordbrug 2522, South
Africa



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HTR2004                                                                    Beijing, CHINA, 2004. 9.



REFERENCES

[1]   S. Ion, D. Nicholls, R. Matzie and D. Matzner: Pebble Bed Modular Reactor: The First
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[2]   G.H. Lohnert: Technical design features and essential safety-related properties of the
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[3]   P. Zanocco, M. Giménez and D. Delmastro: Safety design maps: an early evaluation of
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[4]   W. Bernnat, W. Feltes: Models for reactor physics calculations for HTR pebble bed
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[5]   N. Ben Said: The Impact of Design on the Decay Heat Removal Capabilities of a
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[6]   K. Hoffmann, K.H. Finger, K. Stute: MARS Mass Storage and Retrieval System
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[7]   T. Kindt; H. Haque: Recriticality of the HTR-Modul Power Reactor after Hypothetical
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[8]   Labview, User Manual, National Instruments; January 1998 Edition;
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[9]   K. H. Finger, K. Hoffmann: MARS: Mass Storage Archieving and Retrieval System
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[10] A. Walter, A. Schultz, G. Lohnert: Comparison of Two Models for a Pebble Bed
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     Meeting on HTR Technology, Beijing, INET China September 22-24, 2004
[11] G.H. Lohnert: The ultimate safety of the HTR-MODULE during hypothetical accidents.
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     International Atomic Energy Agency, Vienna (Austria)
      IAEA-TECDOC-757, pp.47-50, 1992
[12] W. Scherer; H. Gerwin, T. Kindt and W. Patscher: Analysis of reactivity and
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     63 (1987)
[13] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High-
     Temperature Gas-Cooled Reactors: Part 3: Loss of Pressure through Friction in
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[14] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High-
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[15] Nuclear Safety Standards Commission (KTA): Reactor Core Design of High-
     Temperature Gas-Cooled Reactors: Part 1: Calculation of Material Properties of
     Helium (June 1978) KTA 3102.1




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