Table Of ContentADVANCED REACTORS:
PHYSICS, DESIGN AND ECONOMICS
Proceedings of the International Conference held
at Atlanta, Georgia, September 8—11, 1974
Edited by
J. M. Kallfelz and R. A. Karam
School of Nuclear Engineering
Georgia Institute of Technology
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Copyright © 1975, American Nuclear Society
All Rights Reserved. No part of this publication may be
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First edition 1975
Library of Congress Cataloging in Publication Data
Main entry under title:
Advanced reactors.
1. Fast reactors — Congresses. 2. Nuclear reactors -
Congresses. 3. Nuclear engineering — Congresses.
I. Kallfelz, J.M., ed. II. Karam, Ratib Abraham,
1934- ed.
TK9203.F3A34 1975 621.48'31 75-4642
ISBN 0-08-019610-1
Printed in Great Britain by A. Wheaton and Co., Exeter
PREFACE
The purpose of the conference was to bring together experts in the fields
of physics, design and economics of advanced reactors, to discuss and exchange
information pertinent to the development of such reactors. The meeting was
quite timely in light of the increased worldwide awareness of limitations on
energy resources, combined with the potential for long term and low cost
energy from nuclear power and the active and large research programs in many
nations for advanced reactor development.
A wide range of topics was covered, including current national programs,
economics of advanced systems, progress in theoretical methods, cross sections
and their impact on design, problems of design and their solutions, and opera
tional experience with specific advanced systems.
The meeting was international in scope, with participation from Austria,
Belgium, Brazil, Canada, France, Germany, Israel, Japan, Spain, the United
Kingdom, the United States and the USSR. A total of sixty papers were pre
sented, twenty-one of which were from abroad.
The conference was sponsored by the following organizations:
ANS Reactor Physics Division
ANS Atlanta Section
ANS Nuclear Fuel Cycle Division
ANS Power Division
School of Nuclear Engineering,
Georgia Institute of Technology
-Southern Interstate Nuclear Board
We would like to acknowledge the contribution of the members of the local
ANS committee and the technical program committee. In particular, we would
like to thank W. Scott Fellows and G.L. Weaver, who devoted considerable time
to the administration of the conference, and to the members of the program
committee who participated in the program review session: Tony Buhl, Walt
Kato, Jack Ohanian, S. Ramchandran, Bert Toppel, Jim Tulenko, and Chuck
Weisbin. Mrs. Phyllis Frost, Mrs. Lydia Geeslin and Ms. Flo Bigham were very
helpful in the preparations for the meeting and the proceedings.
Dr. Lynn Weaver, Director of the School of Nuclear Engineering at Georgia
Tech, was the General Co-Chairman and was responsible for the overall planning
and administration of the meeting. Through him, extensive facilities of
Georgia Tech were made available to the conference at no cost.
We are grateful to the following members of the Student ANS Chapter at
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xiv
Georgia Tech, who provided assistance during the technical sessions: D. Lai,
B. Lancaster, D. Pitts, W. Reupke, D. Weaver, J. Weaver and J. Wohlgemuth.
The majority of the papers are reproduced essentially as they were re
ceived from the authors, with little or no change. We concentrated on editing
some of the papers for which we felt that the grammar was most in need of
revision to eliminate ambiguity. For some of these papers, and for all the
questions and answers from the discussion periods, schedules and deadlines
did not allow confirmation of the edited portions by the authors. For any
errors which many have thus been introduced in the text the editors take sole
responsibility.
In closing we acknowledge the patience and understanding of our families
during our period of involvement with preparations for the meeting and the
publication of the proceedings.
J.M. Kallfelz and R.A. Karam
School of Nuclear Engineering
Georgia Institute of Technology
MEETING ORGANIZATION, ATLANTA ANS SECTION
General Co-Chairmen
Paul F. Schutt, Nuclear Assurance Corp. Lynn E. Weaver, Georgia Tech
Technical Program Co-Chairmen
John M. Kallfelz, Georgia Tech R. A. Karam, Georgia Tech
Local ANS Committee
C. E. Upright, W. Scott Fellows (Finance); R. Whitfield, R. Zimmerman
(Hotel); W. H. Ollinger, G.L. Weaver (Registration); T. E. Byerly,
R. L. Cubitt (Press Relations); G. G. Eichholz, C. D. Strickland
(Luncheon); W. W. Graham, III, F. C. Apple (Ladies Program)
2
FRENCH PROGRAM OF ADVANCED REACTORS
C. Clouet d'Orval, J. Brisbois, J. P. Schwartz, J. C. Mougniot
Commissariat a l'Energie Atomique
29, rue de la Federation
Paris, France
INTRODUCTION
At a time when an important programme (_1) of commercial plants is near
completion, one wonders whether an effort of research and development is
still necessary to further develop advanced technologies.
There are two different kinds of actions in that field. First of all,
there are those actions which are concerned with technologies very different
from existing ones, but where at least one demonstration plant exists today,
excluding for instance GCFR and molten salt reactors. Actually, work on
these different systems is relatively small; furthermore, about fifteen
years ago, many concepts of reactors were studied, from organic reactor to
molten plutonium reactor.
Secondly, there are studies concerning adaptations of existing systems;
it is the case, for example for reactors operating at a given power level,
and for which a new demand appears but at a different power level. It is
the case for light water reactors, for high temperature reactors, and for
sodium cooled reactors. It is also the case for reactor concepts already
developed and which are to be used more efficiently. The adaptation studies
are numerous. The present paper intends to give an idea of what can be done
with systems already used or in the project stage in France.
We shall particularly concentrate on performance, and on core concepts,
of reactors under consideration. To illustrate our purpose, we will speak
only about systems which, in the actual market, have a non negligible oppor
tunity to be carried out to an industrial phase.
I. FAST REACTORS
The French development programme of fast sodium cooled reactors is at
the present time focused on the two operating reactors Rapsodie and Phenix,
and on an important effort of research and development undertaken to design
and build Super Phenix, the construction of which will begin in 1975.
- Rapsodie (Fortissimo), a 40 Mwth reactor, in operation at Cadarache has a
3
full irradiation activity and its very hard spectrum is more and more used
for studying damages to the structures and shield materials.
- Phenix, is a 250 Mwe demonstration plant in operation at Marcoule since
last year.
- Super Phenix, a 1200 Mwe, power station, an integrated reactor, like
Phenix, is the next step of our programme.
1.) Contribution of Phenix
One has to keep in mind the principal steps of the Phenix realization:
- Beginning of construction: 1968
- Test of different components: 1972
- First criticality: August 31, 1973
- First coupling on the grid: December 12, 1973
- Full power: March 12, 1974.
The operation of this reactor is perfectly consistent with its nominal
specifications, and the first measurements have already been described in
several papers.
The very few difficulties during the plant start-up, the excellent
similarity between the computed parameters and those measured give us con
fidence in the future.
We have particularly studied, during the test period neutronic and
thermal behaviour of the reactor, and also performances of the main com
ponents in-core, and out of core, including specifications of fuel.
2.) Extrapolation to higher power ranges (5)
If the demonstration given by Phenix stays, as foreseen, positive, we
look forward to extrapolate to higher power levels. These good results
should allow to begin construction of the Creys - Malville plant (1200 Mwe
Super Phenix Plant) during the year 1975. As it has been done for Phenix,
preliminary studies, then project studies have been completed in order to
give specifications. The preliminary assessments of the experience acquired
from the Phenix construction has shown that the concept and the design are
viable and the technology of components is sufficiently well mastered.
On the fuel side, we are waiting to assess the behaviour of Phenix core
globally and to review in their light what subsequent type of improvements
could be looked for.
In this very consistent forward power move of the fast breeders implan
tation we think we have reached the time where a large size power plant
should be studied and built. Super Phenix is such a reactor and represents
an extrapolation of Phenix. The first 1200 MWe plant is not supposed to
reach competitivity with light water reactors, which is not surprising.
Studies have already been engaged to define the next power plants, which
could be built in a series, and could be quickly competitive mainly if the
uranium prices rise significantly.
4
Approaching commercial phases, it becomes clear that associations and
suitable measures for coordination programs become necessary. A recent
agreement between France, Germany and Italy main producers of electricity:
EDF, RWE, ENEL, has created such association: the first power plant Super
Phenix french type, will be built in France with financial contribution of
51% from EDF, 33% from ENEL and 16% from RWE, the second one, based on
German technique, will have 51% from RWE, 33% from ENEL and 16% from EDF.
3.) Attractiveness of a 450 MWe Reactor (6)
Since studies concerning Phenix and Super Phenix which have a high
priority in our programme will be developed elsewhere in this conference
(2) , (3) , (4_) , we let these authors give details on that subject and focus
this review paper on a particular point. We have examined what power
level we could extrapolate Phenix design up to without major change in com
ponents. The reason is that we think, there is a potential market in
intermediate range of power, medium size electrical network and this reason
has led us to the concept of the 450 MWe Phenix Reactor.
As it has already been described, we will only recall its main options:
- The fuel bundle and the steam generator module are not changed.
- The pumps and intermediate heat-exchangers are directly extrapolated
from the Phenix components. Four secondary loops with eight inter
mediate heat exchangers have been chosen so as to give the lowest
extrapolation factor.
The differences are the following:
- Nominal clad temperature situated between Phenix and Super Phenix
temperatures (630 °C).
- Higher burn-up rate (70,000 MWJ/T) lower fuel and component handling
frequency (shut-down every six months).
- Suppression of in-reactor storage (direct discharge).
- Reactor safety ensured by redundancy of shut-down systems and by a
concrete containment providing anti-missle protection.
The main data of the reactors Phenix 250, Phenix 450 and Super Phenix are
compared in Table 1.
As regards Research and Development our efforts have been continued
with the following aims:
- On the one hand, adaptation to different reorganization schemes for
the fast reactor line, with an optimization study of the breeding
gains and doubling times and also use of carbide fuels.
- On the other hand the study of the present concept extrapolated to
the 2000 MWe range.
5
I PHENIX 250 PHENIX 450 SUPER PHENIX
Gross electrical rating 250 450 1.200
Thermal rating 563 1050 2.990
Core volume 1227 1 2135 1 10.000 1
Length assembly- 4,3 m 4,3 m 5,5 m
Fuel pins per assembly- 217 217 271
Pin outside diameter 6, 6 mm 6,6 mm 8,65 mm
Maximum linear power 430 W/cm 430 W/cm 450 W/cm
Burn-up 50.000 Mtfj/T 70.000 MA/J/T 70.000 Mtfj/T
Maximum total neutron 7,2.1015n/cm2s 7,3 1015n/cm2s 6,2 1015n/cm2s
flux
Breeding ratio 1,12 1 ,15 1,24
! Command basis with 6 9 21
mixed functions
Complementary- shut 0 3 3
down system
Nominal cladding 650/700° C 630/700° C 620/700° C
temperature
Sodium temperature 400/560° C 390/540° C 380/530° C
inle±_/_ 0ntlet_cj2r.fi
Intermediate exchanger 350/550° C 350/530° C 330/510° C
inlet / outler secondary
Primary pumps i 3 4 4
Intermediate exchanger* i 6 8 8
3 4 4
Secondary loops
Primary Na flow 2760 Kg/s 6300 Kg/s 15700 Kg/s
Secondary Na flow 2210 Kg/s 4600 Kg/S 12000 Kg/s
Steam temperature 510° C 495° C 490° C
Inlet H20 temperature 246° C 246° C 235° C
Steam pressure (out) 168 180 184
Table 1
6
4.) Fuel for Fast Breeders
All these reactors described in this paper are loaded with oxide fuel.
As a matter of fact, we are convinced that the commercial demonstration of
fast breeders will be carried out with oxide fuel. However, we are aware
that the breeding ratio of such reactors will not exceed 1.25, correspond
ing to a doubling time longer than 20 years. At that moment, it represents
no hindrance, since the plutonium is coming from thermal reactors. At the
end of year 1990, it would be interesting to have reactors with improved
characteristics. This is the reason for which we started in France a pro
gramme on Carbide fuel which would allow, in about 15 years, to build some
Carbide commercial plants.
II. HIGH TEMPERATURE REACTORS (7_)
In the field of High Temperature reactors, France has concluded agree
ments with Gulf two years ago. For the Nuclear Steam Supply System, we
have access to the techniques in an agreement by which CEA is involved in
an important programme of research and development devoted to this type of
reactor. Other agreements have been concluded for fuel fabrication and in
industrial arrangements. We have almost completed the preparation of an
offer for an HTGR of 1160 MWe, FULTON type. One knows that High Temperature
reactors with helium cooling, besides their application for electrical
power supply with vapor cycle is interesting for other applications which
can be used for different purposes. Therefore, these applications may need
higher temperatures than those which are normally used in the classical
system. We have, at the CEA, started studies in that field for advanced
HTGR and we will give them a general scope on our ideas* in improving temper
atures .
In commercial steam-cycle HTGRs the core consists of graphite hexagonal
blocks containing the fuel. A quarter of the fuel inventory is replaced
each year. Columns of blocks are distributed in refueling regions. Each
region is generally composed of a central column with 6 other columns
around it. A flow gagging takes place at the top of a region in order to
adjust the helium outlet mean temperature. Helium temperature at the heat
exchanger inlet is 750 °C. It is nevertheless possible to reach a tempera
ture higher than 1000 °C by an action in three fields:
- core geometrical and thermodynamical parameters
- core physics
- fuel improvement
1.) Core Geometrical and Thermodynamical Parameters
Without any modification of the fuel itself and its performance, a first
step to a higher helium outlet temperature is obtainable by optimizing:
- the core power density
- the core active height
- the number of fuel holes and channels within a block
- the core inlet temperature.
7
On the Figure 1 we can see the variation of power density as a function
of rise of helium outlet temperature (at the outlet of the channels before
mixing with by-passed and cooling gas) for a core inlet temperature equal to
450 °C. Each curve calculated with the same maximum fuel temperature is
drawn for a fixed core height (6 or 8 blocks) and a given type of block
(block 1160 with 72 channels or blocks FSV with 108 channels).
Figure shows that the influence of power density is small and the main
parameters are the geometry of the block and the height of the core. For
example, at the same power density than the commercial HTGR plant one
(8.4 MW/m3), we can obtain an outlet helium temperature 40 °C higher.
On Figure 2 we can see the influence of helium temperature at the entry
of core.
A 45 °C increase of the outlet temperature is reached for a 100 °C in
crease of the inlet temperature. But the by-pass flow cooling control rods
and other parts of the core increase with the helium inlet temperature.
Then it is reasonable to expect an outlet temperature only 40 °C higher for
temperatures between 350 and 450 °C at the core entry, and 30 °C for inlet
temperatures between 450 and 550 °C. Beyond these values it would be neces
sary to cool control rods by means of a helium by pass at a temperature
lower than the entry one, or to coat control rods with refractory materials.
All these modifications do not change the core nor fuel concept. It
can be seen that 900 °C is achievable. This temperature level is attractive
for helium turbine cycle and many chemical processes. Technological prob
lems can be solved without a significant R and D. For the coal gasification
for instance, it would be nevertheless interesting to reach a temperature
between 900 °C and 1000 °C. This is the second step.
2.) Physical and Conceptual Arrangement of the Core
In this step a modeling of axial and radial power distribution in the
core is proposed.
At the region periphery, a higher flux is found. This is due to dif
ferent ages of neighboring regions or to flux fall around control rods.
This also leads to a high tilt factor influence of which is represented on
Figure 3. Tilt factor is defined as the ratio of local power over mean
power in a region. The dashed curve is drawn with a tilt factor equal to
1.25 and the continuous curve with tilt factor equal to 1.65. These curves
show a difference of 100 °C between outlet temperatures.
In order to reduce the tilt factor, several means are studied:
- a more uniform distribution of control rods
- a use of Pu cycle with refueling of the whole core.
Thus, all fuel columns are equally old, and control rods are not very much
moved because the reactivity of the core is stable during almost all the
life of the core.
- a shuffling of all columns each year. The bottom blocks are removed
