CA1267446A - Nuclear reactor and method of operating same - Google Patents
Nuclear reactor and method of operating sameInfo
- Publication number
- CA1267446A CA1267446A CA000491578A CA491578A CA1267446A CA 1267446 A CA1267446 A CA 1267446A CA 000491578 A CA000491578 A CA 000491578A CA 491578 A CA491578 A CA 491578A CA 1267446 A CA1267446 A CA 1267446A
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/18—Emergency cooling arrangements; Removing shut-down heat
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C7/00—Control of nuclear reaction
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Chemical & Material Sciences (AREA)
- Chemical Kinetics & Catalysis (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
11670/CA/Dr.v.B/Scha/67 Inventor, Post Office Address and Citizenship Hans Volker KLAPDOR, Forstquelle 4, Heidelberg / F.R. Germany Citizen of the Federal Republic of Germany Josef METZINGER, Schwanheimer Str. 16c, Eberbach / F.R. Germany Citizen of the Federal Republic of Germany Invention: Nuclear Reactor and Method of Operating same Abstract of the Disclosure A nuclear reactor or nuclear power plant, as a light water reactor (LWR), including a high-converting LWR, a heavy water reactor, a gas cooled reactor, a high temperature reactor, a fast breeder, with given maximum and effective output power (installed or nominal capacity) and given cooling capacity of the emergency cooling system is safely operated with increased maximum or effective power output which exceeds the values permitted by present nuclear standard and safety regulations. Further, a method for design and construction of a nuclear reactor with reduced cost of the emergency cooling system which is more economical than present methods are disclosed. The improvement is based on a new and more precise way of predicting the decay heat produced in the core of a nuclear reactor after shutdown and makes use of new, more reliable data which show that the decay heat of nuclear reactors is overestimated by present nuclear standards and nuclear safety regulations such as, e.g., ANSI/ANS-5.1-1979 (reaffirmed 13 July 1985 for further 5 years as ANSI/ANS-5.1-1985) and DIN 25463.
Description
7~1~1¢
SPECIFICATION
The present invention rela-tes generally to the production of heat energy by nuclear fission, and more specifically to nuclear reactors, nuclear power plants, procedures for designing and constructing the emergency cooling system of a nuclear reactor.
.
Background of the In~ention The relation between -the maximum effective output power (ins-talled or nominal capacity) of a nuclear reactor and the cooling capacity of its emergency cooling system is essentially determined by the so-called decay heat, i.e. the heat produced by radioactive decay of the fission products after shut-down of the fissioning process.
The removal of the decay heat is most critical, particularly for light water reactors (LWR), in the first minutes after shutdown.
Immediately after shutdown the decay heat corresponds to about 7%
of the operating power of the reactor before shut-down. In case of a pressurized water reactor (PWR), without cooling, the decay heat would cause melting of the metal jackets enclosing the fuel pellets within 50 seconds after shut-down from nominal power.
Typically, the heat produced by a LWR in the first 200 seconds after shutdown determines the necessary maximum cooling capacity of the emergency cooling system of the reactor.
SPECIFICATION
The present invention rela-tes generally to the production of heat energy by nuclear fission, and more specifically to nuclear reactors, nuclear power plants, procedures for designing and constructing the emergency cooling system of a nuclear reactor.
.
Background of the In~ention The relation between -the maximum effective output power (ins-talled or nominal capacity) of a nuclear reactor and the cooling capacity of its emergency cooling system is essentially determined by the so-called decay heat, i.e. the heat produced by radioactive decay of the fission products after shut-down of the fissioning process.
The removal of the decay heat is most critical, particularly for light water reactors (LWR), in the first minutes after shutdown.
Immediately after shutdown the decay heat corresponds to about 7%
of the operating power of the reactor before shut-down. In case of a pressurized water reactor (PWR), without cooling, the decay heat would cause melting of the metal jackets enclosing the fuel pellets within 50 seconds after shut-down from nominal power.
Typically, the heat produced by a LWR in the first 200 seconds after shutdown determines the necessary maximum cooling capacity of the emergency cooling system of the reactor.
- 2 -, . . .
. ' ' :
.
~Z~ 6 The decay heat is essentially comprised by three portions origina-ting from 1. the decay of the fission products, 2- the decay of the elements, like 239u, 239Np and further actinides, which are formed by neutron capture by the nuclear fuel,
. ' ' :
.
~Z~ 6 The decay heat is essentially comprised by three portions origina-ting from 1. the decay of the fission products, 2- the decay of the elements, like 239u, 239Np and further actinides, which are formed by neutron capture by the nuclear fuel,
3. the decay of isotopes produced by neutron capture by the fission products.
The presently valid nuclear standards, as the ANS standard ANST/ANS-5.1-1979/198S and the DIN standard DIN 25463 (July 1982), which estimate the decay heat and basing on this prescribe the minimum permitted cooling capacity of the emergency cooling system in reldtion to the maximum operating power (installed capacity or nominal output power) of the reactor in normal operation or, in other words, prescribe the maximum permitted operating power of a power plant in normal operation with a given cooling capacity of the emergency cooling system, are based on relatively uncertain ass~mptions. The experimental information available at present is incomplete, as discussed in the report on the "Invited Paper presented at International Conference on Nuclear Power Plant Aging, Availability Factor and Reliability Analysis", San Diego, California, 8 - 12 July 1985, Max-Planck-Institut fur Kernphysik, Heidelberg, F. R. Germany, MPI H-1985-V14, and is not adapted for a preclse prediction of the decay heat, in particular for the ~irst minutes after shut-down. It also was not possible to * copy attach~d .
~ '. , ,' ' '' ' ~ ' ' ., '' ~
reliably theoretically predict the decay heat, as also discussed in the above report, in particular for the first minutes after shutdown, because rough approximations had to be made and particularly only incorrect descriptions existed of the beta decay of the involved nuclides.
The consequence of the latter is - this is found as result of new data based on a more precise theoretical method of calculating the nuclear ~ decay of the various fission products, which form the basis of the invention to be described - a considerable overdimensioning of reactor emergency cooling systems up to now.
Summary of the In~entio~
An object of the present invention is to provide a method for operating nuclear reactors and power plants which is more . economical than the methods of the prior art.
A further object of the invention is to provide nuclear reactors, which differ advantageously from the known nuclear reactors by ~ higher efficiency and maximum power output at given cost of installation, or by lower costs of installation at given maximum output power.
The present invention is based on new and reliable data of the amount of the decay heat - including all of the three portions thereof mentioned above. We have Found, that the total decay heat
The presently valid nuclear standards, as the ANS standard ANST/ANS-5.1-1979/198S and the DIN standard DIN 25463 (July 1982), which estimate the decay heat and basing on this prescribe the minimum permitted cooling capacity of the emergency cooling system in reldtion to the maximum operating power (installed capacity or nominal output power) of the reactor in normal operation or, in other words, prescribe the maximum permitted operating power of a power plant in normal operation with a given cooling capacity of the emergency cooling system, are based on relatively uncertain ass~mptions. The experimental information available at present is incomplete, as discussed in the report on the "Invited Paper presented at International Conference on Nuclear Power Plant Aging, Availability Factor and Reliability Analysis", San Diego, California, 8 - 12 July 1985, Max-Planck-Institut fur Kernphysik, Heidelberg, F. R. Germany, MPI H-1985-V14, and is not adapted for a preclse prediction of the decay heat, in particular for the ~irst minutes after shut-down. It also was not possible to * copy attach~d .
~ '. , ,' ' '' ' ~ ' ' ., '' ~
reliably theoretically predict the decay heat, as also discussed in the above report, in particular for the first minutes after shutdown, because rough approximations had to be made and particularly only incorrect descriptions existed of the beta decay of the involved nuclides.
The consequence of the latter is - this is found as result of new data based on a more precise theoretical method of calculating the nuclear ~ decay of the various fission products, which form the basis of the invention to be described - a considerable overdimensioning of reactor emergency cooling systems up to now.
Summary of the In~entio~
An object of the present invention is to provide a method for operating nuclear reactors and power plants which is more . economical than the methods of the prior art.
A further object of the invention is to provide nuclear reactors, which differ advantageously from the known nuclear reactors by ~ higher efficiency and maximum power output at given cost of installation, or by lower costs of installation at given maximum output power.
The present invention is based on new and reliable data of the amount of the decay heat - including all of the three portions thereof mentioned above. We have Found, that the total decay heat
- 4 -. . . .
is smaller than -that, which has been assumed on -the basis of the measurements and calculations available before the invention.
This allows to operate a nuclear reactor, which has a cooling sys-tem of a predetermined cooling power capacity, with higher effective and maximum output power (nominal power), particular also in normal continuous operation. This is of particular interest for existing LWR's, heavy water reactors and gas cooled reactors used in large scale for the production of electric power, and can be achieved by a simple modification of the control and safety systems of the reactor.
Further, the emergency cooling systems of a nuclear reactor of given maximum output power (installed capacity of nominal output power) can be designed smaller than according to the teaching of the prior art.
The latter is specifically of considerable importance for LWR's, further for heavy water reactors and gas cooled reactors. The same is true for high-converting LWR's, high tem-perature reactors and fast breeders, for which only very scarce experimental decay heat data exist up to now.
2~ In accordance with a broad aspect of the invention there is provided a method of operating a nuclear reactor and power plant, which is designed for a given maximum effective power output (installed capacity) and which has an emergency cooling system having a given cooling capacity for removal of decay heat, where the decay heat-removal capacity is a given percentage of the maxi-mum admissible effective power (installed capacity~, characteri~ed ~i7~6 238~9-22 in that said percentage ls smaller than the value re~uired by the nuclear standards and regu1.ations valid in 1.984.
In accordance with another broad aspect oE the inven-tion there is pxovided a nuclear reactor comprising a control system which limits the maximum permitted effective (nominal) out-put power in continuous operation and an emergency cooling system having a cooling capacity proportioned to said nominal power . output, characterized in that said control system is devised so that the maximum permitted effective power output of the reactor is up to at least 5% larger than that, which follows from the cooling capacity of the emergency cool:ing system according to the nuclear standards and safety regulations valid in 1984.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor comprising a control system, which limits the maximum nominal e~fective power output in normal operation, and an emergency cooling system having a cooling capacity proportioned to said nominal power output, char-acterized in that said control system is designed for a maximum and effective power output in normal operation, which exceeds the maximum and effective power output permitted by the standard . DIN 25436.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor comprising a control system, which limits the maximum nominal effective power outpu-t in normal operation~ and an emergency cooling system havi.ng a cooling capacity proportioned to said nominal power output, characterized - 5a -- ~: ' ~7~
238~9-22 in that said control system is designed for a maximum and effec-tive power output in normal operation, which exceeds the maximum and effective power ou-tput ptermitted by the ANS standard ANSI/ANS-5.1-1979/1985.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor as defined above, char-acterized in that, the maximum and effective power output is by at most 5~ larger than the value according to said standard.
In the following preferred embodiments of the invention will be explained in more detail with reference to the drawings.
Short Descrip-tion of the drawings:
is smaller than -that, which has been assumed on -the basis of the measurements and calculations available before the invention.
This allows to operate a nuclear reactor, which has a cooling sys-tem of a predetermined cooling power capacity, with higher effective and maximum output power (nominal power), particular also in normal continuous operation. This is of particular interest for existing LWR's, heavy water reactors and gas cooled reactors used in large scale for the production of electric power, and can be achieved by a simple modification of the control and safety systems of the reactor.
Further, the emergency cooling systems of a nuclear reactor of given maximum output power (installed capacity of nominal output power) can be designed smaller than according to the teaching of the prior art.
The latter is specifically of considerable importance for LWR's, further for heavy water reactors and gas cooled reactors. The same is true for high-converting LWR's, high tem-perature reactors and fast breeders, for which only very scarce experimental decay heat data exist up to now.
2~ In accordance with a broad aspect of the invention there is provided a method of operating a nuclear reactor and power plant, which is designed for a given maximum effective power output (installed capacity) and which has an emergency cooling system having a given cooling capacity for removal of decay heat, where the decay heat-removal capacity is a given percentage of the maxi-mum admissible effective power (installed capacity~, characteri~ed ~i7~6 238~9-22 in that said percentage ls smaller than the value re~uired by the nuclear standards and regu1.ations valid in 1.984.
In accordance with another broad aspect oE the inven-tion there is pxovided a nuclear reactor comprising a control system which limits the maximum permitted effective (nominal) out-put power in continuous operation and an emergency cooling system having a cooling capacity proportioned to said nominal power . output, characterized in that said control system is devised so that the maximum permitted effective power output of the reactor is up to at least 5% larger than that, which follows from the cooling capacity of the emergency cool:ing system according to the nuclear standards and safety regulations valid in 1984.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor comprising a control system, which limits the maximum nominal e~fective power output in normal operation, and an emergency cooling system having a cooling capacity proportioned to said nominal power output, char-acterized in that said control system is designed for a maximum and effective power output in normal operation, which exceeds the maximum and effective power output permitted by the standard . DIN 25436.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor comprising a control system, which limits the maximum nominal effective power outpu-t in normal operation~ and an emergency cooling system havi.ng a cooling capacity proportioned to said nominal power output, characterized - 5a -- ~: ' ~7~
238~9-22 in that said control system is designed for a maximum and effec-tive power output in normal operation, which exceeds the maximum and effective power ou-tput ptermitted by the ANS standard ANSI/ANS-5.1-1979/1985.
In accordance with another broad aspect of the inven-tion there is provided a nuclear reactor as defined above, char-acterized in that, the maximum and effective power output is by at most 5~ larger than the value according to said standard.
In the following preferred embodiments of the invention will be explained in more detail with reference to the drawings.
Short Descrip-tion of the drawings:
- 5~ -L~
4~6 Fig. 1 shows the calculated so-called burst-function for fast fission of ~38U.
Fig. 2a shows in diagrammatic form the overestimate by the ANS
(dashed line) and the DIN (solid line) standards of the total decay heat produced from the time of shutdown (t-O) of a typical pressurized water reactor (PWR) till time t after shutdown, in percent of the newly determined total decay heat ETHOR on which the invention is based. The diagrams of Fig. 2a correspond to a PWR with a burn-up of 38 megawatt-days (MWd) per kg of nuclear fuel, the latter being assumed as consisting of uranium enriched to 3.4% in - 235U at the time of starting the reactor.
Fig. 2b and 2c are diagrams corresponding to Fig. 2a, but for a burn-up of 19 MWd/kg and 53~4 MWd/kg, respectively.
Fig. 3 are diagrams for the case of the PWR of Fig. 2a, showing the situation relative to the ANS prediction for those parts of the decay heat originating from neutron capture by the fission products (solid line) and by the actinides (dashed line).
Fig. 4 are diagrams corresponding to those of Fig. 3, but referring to the DIN standard.
~2~
Fig. 5 are diagrams corresponding to Fig. 2a, but for a boiling water reactor (BWR) and a burn-up of 23 MWd/kg of the nuclear fuel mentioned with reference to Fig. 2a;
Fig. 6 are diagrams corresponding to Fig. 2a, but for a high-con-verting LWR and a burn-up of 38.4 MWd/kg of nuclear fuel consisting of 0 2 % 235u 88.65% 238u,
4~6 Fig. 1 shows the calculated so-called burst-function for fast fission of ~38U.
Fig. 2a shows in diagrammatic form the overestimate by the ANS
(dashed line) and the DIN (solid line) standards of the total decay heat produced from the time of shutdown (t-O) of a typical pressurized water reactor (PWR) till time t after shutdown, in percent of the newly determined total decay heat ETHOR on which the invention is based. The diagrams of Fig. 2a correspond to a PWR with a burn-up of 38 megawatt-days (MWd) per kg of nuclear fuel, the latter being assumed as consisting of uranium enriched to 3.4% in - 235U at the time of starting the reactor.
Fig. 2b and 2c are diagrams corresponding to Fig. 2a, but for a burn-up of 19 MWd/kg and 53~4 MWd/kg, respectively.
Fig. 3 are diagrams for the case of the PWR of Fig. 2a, showing the situation relative to the ANS prediction for those parts of the decay heat originating from neutron capture by the fission products (solid line) and by the actinides (dashed line).
Fig. 4 are diagrams corresponding to those of Fig. 3, but referring to the DIN standard.
~2~
Fig. 5 are diagrams corresponding to Fig. 2a, but for a boiling water reactor (BWR) and a burn-up of 23 MWd/kg of the nuclear fuel mentioned with reference to Fig. 2a;
Fig. 6 are diagrams corresponding to Fig. 2a, but for a high-con-verting LWR and a burn-up of 38.4 MWd/kg of nuclear fuel consisting of 0 2 % 235u 88.65% 238u,
6 44% 239PU
2 96% 240pU
1.06% 241pU
0.69% Pu;
Fig. 7 are diagrams corresponding to Fig. 2a, but for a heavy water reactor of the `'CANDU" type and a burn-up of 15.1 MWd/kg of nuclear fuel consisting of natural uranium.
~. .
, : ' :
~6~
Fig. 8 comprising parts 1 to 18 is a TABLE comprising a list of data for computing the decay heat by means of equations (3) and (4) explained in the Appendix, wherein is:
Z the nuclear charge of the isotope;
A the mass number of the isotope;
K denotes, whether the isotope decays from the ground state (G) or from an isomeric state (I);
~tot the total decay rate of isotope Z,A,K, per second;
P~-(G) the probability of decay by ~ emission finally feeding the ground state of the daughter nucleus;
P~ ) the probability of decay by r~ emission finally ~eeding an isomeric state of the daughter nucleus;
P~-~(G) the probability of decay by r~ emission finally feeding the ground state of the daughter nucleus;
P~+(I) the probability of decay by ~+ emission finally feeding an isomeric state of the daughter nucleus;
P~ the probability of ~ decay;
Py the probability of y decay of an isomeric state;
PSF the probability of decay by spontaneous fission;
Pln the probability of decay by ~ delayed one-neutron emission;
P2n the probability o-f decay by ~ delayed two-neutron emission;
ETotal the total reco~erable decay energy (decay heat dissipated in the reactor core), produced by the decay of isotope Z, A~ K, in MeV.
~26~ 6 It is known to those scilled in the art, how to calculate the isotopic content of an arbitrary nuclear reactor at any time during reactor operation and after shutdown as a function of the reactor operation parameters and reactor running time (and also as a special case the isotopic distribution in a sample of fissile material as a function of time after a short-time irradiation of the sample by neutrons). This approach is used to calculate the isotopic inventory of the reactor under consideration. The decay heat produced at time t after shutdown (or irradiation is calculated by integration over the inventory at time t and over the energy contributions from the individual isotopes. Further integration over the time t from shutdown yields the total decay heat produced until time t after shutdown.
The invention is based on new and more reliable data of the decay heat contributed by isotopes, for which no sufficiently precise experimental data were available. These heat energy contributions are listed in the TABLE of Fig. 8 and were obtained by using a microscopic nuclear model, known in principle per se for other purposes, in modified and extended form for the calculation of the decay heat. Thus, the decay heat data for the above calculations are taken from the TABLE in Fig. 8, which comprises new and more reliable data.
~2& f ~
In Fig. 1 as a specific example oF the new information obtainable with -the new data comprised in the TABLE, the so-called burst-function is shown for the fission of 238U by fast neutrons ("fast fission"), i.e. the decay heat per fission event multiplied by the time t after a short-time irradiation of 238U with fast neutrons, as function of time t. The solid curve shows the portion of the decay heat produced by beta radiation, the dashed line the protion produced by gamma radiation. It should be pointed out, that known experimentally obtained burst functions of the above type are not suitable for calculating the decay heat of a reactor, because they do not comprise information on the effects of neutron capture by the fission products and actinides..
Fig. 2a shows that the amount of the total decay heat of a reactor, determined in this new precise way, is - for example for a PWR with an at present usual burn-up of 38 MWdtkg - in the order of 6 to 8% less than the amount assumed up to now, which is the basis of the present saFety standards. The same applies for smaller and, more important, for larger burn-up values, as shown in Figs. 2b and 2c, respectively, and other types of nuclear reactors, as shown in the examples of Figs. S to 7.
Thus, the new values of the amount of the decay heat allow to increase the effective and maximum power output oF a reactor comprising an emergency cooling system of a given cooling capacity beyond the "installed" or nominal capacity" or originally set nominal output power or, when devising a new reactor, to dimension , .
:
~2~E7~
the emergency cooling system smaller than assumed to be necessary before, without impairing the safety.
Already an increase of the effective and maximum power output of only 1% yields very considerable economic~benefits, correspon-dingly larger benefits follow from an increase by 2%, 3%, 4%, 5%, etc.
Figs. 3 and 4 show by way of example that the ANS and DIN
standards make considerably erroneous estimates of the various contributions to the decay heat and that the standards do not always overestimate these partial contributions but sometimes underestimate them - so in the case of the ANS standard the contribution to the decay heat from neutron capture by the fission products. The situation is more or less similar for the other contributions to the decay heat. Only the precise determination of all these portions of or contributions to the decay heat mentioned above, on which this invention is based, allows to reduce the decay heat-removal capacity of the emergency coo1ing system for given installed capacity of a reactor to the optimum value, or to operate, at given emergency cooling capacity, the reactor with correspondingly larger maximum and effective power output.
Figs. 2 to 4 show, that for a PWR with typical burn-up, the reduction of the decay heat is about 6 to 8%. Since, after shutdown of the reactor~ the emergency cooling system has to remove also the specific heat stored in -the reactor core, which in i~3 :
~2~i74~6 the critical time tfirst minutes) is of the same order as the decay heat, the maximum and effective power output of the nuclear power plant can be increased by about 3 to 4%, depending on the operation parameters in appropriate instances by 5% or more.
For other operating parameters than that of Figs. 2 to 4, and for other reactor types in general somewhat different values are obtained, as shown by the examples in Fig. 5 to 7, so that the percentage of the maximum permissible reactor power output, which has to be held available as decay heat-removal capacity of the emergency cooling system, can be up to 10% less, and in appropriate instances even more less, than required by the present standards and safety regulations. In other specific cases a reduction of 8% or 6% or about 5% may be appropriate.
It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptions and the same are intended to be comprehended within the meaning nnd range of equivalents of the appended claims.
44~i APPENDIX
. . .
The calculational procedure to determine the isotopic inventory of a nuclear reactor at any time during operation and after shutdown and to calculate the decay heat as Eunction of running time of the reactor and of time after reactor shutdown is similar to that described by A. Tobias in Prog. Nucl. Energy 5 1 - 93, 1980.
The calculations should preferably be performed using an extended range of magnitude from 10 308 to 1038.
More specifically, for the calculation of the isotopic inventory at an arbitrary point in the core of a reactor linear decay chains of the following form have to be set up and solved:
Nl N2--~N3--~N4 '' ~ Ni i~ .,. (1) wherein Ni denotes the number of the nuclei of isotope i specified by nuclear charge Z, atomic mass A, and K as defined with reference to Figure 8. For any isotope 1 in any decay chain is the rate of change of Ni at time t:
- (Ni) =-~i N + ~iF f ~ Ni-l (2) ~D
wherein ,\jtot is the total decay rate per second, which is listed for each isotope l(Z,A,K) in the TABLE of Fig. 8, and ~tot ~B (G) ~ ~B (I) ~ ;~B (G) + ~B (I) + ;~SF ~ + ,,~ln+ ~2n+ an p + ~f (p;
l is one of the partial decay rates ~B (G) ~R (I~ ~.B (G) ~B (I) ;~ .S. ~ln ~2n ~n ~p with which the (i-l)th isotope contributes to the decay chain.
~x = ~tot px ;
wherein x is any of B (G), B (I), ~ (G), B (I), ~, ~, SF, ln, 2n;
pjx are the values listed in the TABLE for the isotope i(Z,A,K), F is the fission rate per second, is the fission yield of isotope i, ;s the neutron flux per cm2 and second and ~j~ and of are the cross sections (in cm2) for neutron capture without and with successive fission, respect;vely.
- , .
The Fission yield values are taken from a standard nuclear data file, such as ENDF/B-IV, V or VI, Brookhaven, National Lab. When settingup the decay chains, "self-production" of an isotope is to be avoided.
.
~lZ~'4g~6 For typical reactor operation conditions for each unit cell in the reactor a system of equations of the Form of eq. (2) are to be solved for about 1400 isotopes. For each member of a decay chain an analytical solution is used which has the following form:
Ni~t~ k j r ~-r ~ ~r IA~
J~r i wherein ~,tot _ Atot) ~ 1 for ~ 5 m.
The starting abundances NO and the neutron capture cross sections are reactor specific and have to be provided by the operator of the specific reactor under consideration.
During the time of operation of the reactor the relative composi-tion of the fissile material in the core changes according to eq.
(3). Thus the corresponding change of the parameters F and ~j has to be taken into account. For this purpose, the time of operation is divided into intervals with constant fission rates and fission ': ' . ' ' - . . . .
' :lZ~ 4~
yields. This means approximating the real function shape of the reactor power output by a step function ("power histogram").
For typical reactor conditions, the number of time steps should be sufficiently large, so that the quantities to be calculated do not depend any more (within the required accuracy) on the number o-f time intervals. The time intervals must be small compared to the total decay rates of the isotopes involved.
The calculation of the decay heat H at time t is performed using the results of eq. (3) in eq. (4).
Hlt) = ~ ~i Ni(t) ETotal,i The calculation of the integral decay heat from shutdown up to time t is performed by numerical integration of eq. (4) with the Gaussian integration method. The total decay energ;es ETotal j of the individual isotopes are taken from the TABLE of Fig. 8.
.
2 96% 240pU
1.06% 241pU
0.69% Pu;
Fig. 7 are diagrams corresponding to Fig. 2a, but for a heavy water reactor of the `'CANDU" type and a burn-up of 15.1 MWd/kg of nuclear fuel consisting of natural uranium.
~. .
, : ' :
~6~
Fig. 8 comprising parts 1 to 18 is a TABLE comprising a list of data for computing the decay heat by means of equations (3) and (4) explained in the Appendix, wherein is:
Z the nuclear charge of the isotope;
A the mass number of the isotope;
K denotes, whether the isotope decays from the ground state (G) or from an isomeric state (I);
~tot the total decay rate of isotope Z,A,K, per second;
P~-(G) the probability of decay by ~ emission finally feeding the ground state of the daughter nucleus;
P~ ) the probability of decay by r~ emission finally ~eeding an isomeric state of the daughter nucleus;
P~-~(G) the probability of decay by r~ emission finally feeding the ground state of the daughter nucleus;
P~+(I) the probability of decay by ~+ emission finally feeding an isomeric state of the daughter nucleus;
P~ the probability of ~ decay;
Py the probability of y decay of an isomeric state;
PSF the probability of decay by spontaneous fission;
Pln the probability of decay by ~ delayed one-neutron emission;
P2n the probability o-f decay by ~ delayed two-neutron emission;
ETotal the total reco~erable decay energy (decay heat dissipated in the reactor core), produced by the decay of isotope Z, A~ K, in MeV.
~26~ 6 It is known to those scilled in the art, how to calculate the isotopic content of an arbitrary nuclear reactor at any time during reactor operation and after shutdown as a function of the reactor operation parameters and reactor running time (and also as a special case the isotopic distribution in a sample of fissile material as a function of time after a short-time irradiation of the sample by neutrons). This approach is used to calculate the isotopic inventory of the reactor under consideration. The decay heat produced at time t after shutdown (or irradiation is calculated by integration over the inventory at time t and over the energy contributions from the individual isotopes. Further integration over the time t from shutdown yields the total decay heat produced until time t after shutdown.
The invention is based on new and more reliable data of the decay heat contributed by isotopes, for which no sufficiently precise experimental data were available. These heat energy contributions are listed in the TABLE of Fig. 8 and were obtained by using a microscopic nuclear model, known in principle per se for other purposes, in modified and extended form for the calculation of the decay heat. Thus, the decay heat data for the above calculations are taken from the TABLE in Fig. 8, which comprises new and more reliable data.
~2& f ~
In Fig. 1 as a specific example oF the new information obtainable with -the new data comprised in the TABLE, the so-called burst-function is shown for the fission of 238U by fast neutrons ("fast fission"), i.e. the decay heat per fission event multiplied by the time t after a short-time irradiation of 238U with fast neutrons, as function of time t. The solid curve shows the portion of the decay heat produced by beta radiation, the dashed line the protion produced by gamma radiation. It should be pointed out, that known experimentally obtained burst functions of the above type are not suitable for calculating the decay heat of a reactor, because they do not comprise information on the effects of neutron capture by the fission products and actinides..
Fig. 2a shows that the amount of the total decay heat of a reactor, determined in this new precise way, is - for example for a PWR with an at present usual burn-up of 38 MWdtkg - in the order of 6 to 8% less than the amount assumed up to now, which is the basis of the present saFety standards. The same applies for smaller and, more important, for larger burn-up values, as shown in Figs. 2b and 2c, respectively, and other types of nuclear reactors, as shown in the examples of Figs. S to 7.
Thus, the new values of the amount of the decay heat allow to increase the effective and maximum power output oF a reactor comprising an emergency cooling system of a given cooling capacity beyond the "installed" or nominal capacity" or originally set nominal output power or, when devising a new reactor, to dimension , .
:
~2~E7~
the emergency cooling system smaller than assumed to be necessary before, without impairing the safety.
Already an increase of the effective and maximum power output of only 1% yields very considerable economic~benefits, correspon-dingly larger benefits follow from an increase by 2%, 3%, 4%, 5%, etc.
Figs. 3 and 4 show by way of example that the ANS and DIN
standards make considerably erroneous estimates of the various contributions to the decay heat and that the standards do not always overestimate these partial contributions but sometimes underestimate them - so in the case of the ANS standard the contribution to the decay heat from neutron capture by the fission products. The situation is more or less similar for the other contributions to the decay heat. Only the precise determination of all these portions of or contributions to the decay heat mentioned above, on which this invention is based, allows to reduce the decay heat-removal capacity of the emergency coo1ing system for given installed capacity of a reactor to the optimum value, or to operate, at given emergency cooling capacity, the reactor with correspondingly larger maximum and effective power output.
Figs. 2 to 4 show, that for a PWR with typical burn-up, the reduction of the decay heat is about 6 to 8%. Since, after shutdown of the reactor~ the emergency cooling system has to remove also the specific heat stored in -the reactor core, which in i~3 :
~2~i74~6 the critical time tfirst minutes) is of the same order as the decay heat, the maximum and effective power output of the nuclear power plant can be increased by about 3 to 4%, depending on the operation parameters in appropriate instances by 5% or more.
For other operating parameters than that of Figs. 2 to 4, and for other reactor types in general somewhat different values are obtained, as shown by the examples in Fig. 5 to 7, so that the percentage of the maximum permissible reactor power output, which has to be held available as decay heat-removal capacity of the emergency cooling system, can be up to 10% less, and in appropriate instances even more less, than required by the present standards and safety regulations. In other specific cases a reduction of 8% or 6% or about 5% may be appropriate.
It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptions and the same are intended to be comprehended within the meaning nnd range of equivalents of the appended claims.
44~i APPENDIX
. . .
The calculational procedure to determine the isotopic inventory of a nuclear reactor at any time during operation and after shutdown and to calculate the decay heat as Eunction of running time of the reactor and of time after reactor shutdown is similar to that described by A. Tobias in Prog. Nucl. Energy 5 1 - 93, 1980.
The calculations should preferably be performed using an extended range of magnitude from 10 308 to 1038.
More specifically, for the calculation of the isotopic inventory at an arbitrary point in the core of a reactor linear decay chains of the following form have to be set up and solved:
Nl N2--~N3--~N4 '' ~ Ni i~ .,. (1) wherein Ni denotes the number of the nuclei of isotope i specified by nuclear charge Z, atomic mass A, and K as defined with reference to Figure 8. For any isotope 1 in any decay chain is the rate of change of Ni at time t:
- (Ni) =-~i N + ~iF f ~ Ni-l (2) ~D
wherein ,\jtot is the total decay rate per second, which is listed for each isotope l(Z,A,K) in the TABLE of Fig. 8, and ~tot ~B (G) ~ ~B (I) ~ ;~B (G) + ~B (I) + ;~SF ~ + ,,~ln+ ~2n+ an p + ~f (p;
l is one of the partial decay rates ~B (G) ~R (I~ ~.B (G) ~B (I) ;~ .S. ~ln ~2n ~n ~p with which the (i-l)th isotope contributes to the decay chain.
~x = ~tot px ;
wherein x is any of B (G), B (I), ~ (G), B (I), ~, ~, SF, ln, 2n;
pjx are the values listed in the TABLE for the isotope i(Z,A,K), F is the fission rate per second, is the fission yield of isotope i, ;s the neutron flux per cm2 and second and ~j~ and of are the cross sections (in cm2) for neutron capture without and with successive fission, respect;vely.
- , .
The Fission yield values are taken from a standard nuclear data file, such as ENDF/B-IV, V or VI, Brookhaven, National Lab. When settingup the decay chains, "self-production" of an isotope is to be avoided.
.
~lZ~'4g~6 For typical reactor operation conditions for each unit cell in the reactor a system of equations of the Form of eq. (2) are to be solved for about 1400 isotopes. For each member of a decay chain an analytical solution is used which has the following form:
Ni~t~ k j r ~-r ~ ~r IA~
J~r i wherein ~,tot _ Atot) ~ 1 for ~ 5 m.
The starting abundances NO and the neutron capture cross sections are reactor specific and have to be provided by the operator of the specific reactor under consideration.
During the time of operation of the reactor the relative composi-tion of the fissile material in the core changes according to eq.
(3). Thus the corresponding change of the parameters F and ~j has to be taken into account. For this purpose, the time of operation is divided into intervals with constant fission rates and fission ': ' . ' ' - . . . .
' :lZ~ 4~
yields. This means approximating the real function shape of the reactor power output by a step function ("power histogram").
For typical reactor conditions, the number of time steps should be sufficiently large, so that the quantities to be calculated do not depend any more (within the required accuracy) on the number o-f time intervals. The time intervals must be small compared to the total decay rates of the isotopes involved.
The calculation of the decay heat H at time t is performed using the results of eq. (3) in eq. (4).
Hlt) = ~ ~i Ni(t) ETotal,i The calculation of the integral decay heat from shutdown up to time t is performed by numerical integration of eq. (4) with the Gaussian integration method. The total decay energ;es ETotal j of the individual isotopes are taken from the TABLE of Fig. 8.
.
Claims (30)
1. A method of operating a nuclear reactor and power plant, which is designed for a given maximum effective power output (installed capacity) and which has an emergency cooling system having a given cooling capacity for removal of decay heat, where the decay heat-removal capacity is a given percentage of the maximum admissible effective power (installed capacity), characterized in that said percentage is smaller than the value required by the nuclear standards and regulations valid in 1984.
2. The method as claimed in claim 1 characterized in that said percentage is at least 1% smaller than the value prescribed by said standards and safety regulations.
3. The method as claimed in claim 1, characterized in that said percentage is smaller by at least 2% than the value prescribed by said standards and safety regulations.
4. The method as claimed in claim 1, characterized in that said percentage is at least 3% smaller than the value prescribed by said standards and safety regulations.
5. The method as claimed in claim 1 characterized in that said percentage is no more than 10% smaller than the value prescribed by said standards and safety regulations.
6. The method as claimed in claim 1, characterized in that said percentage is no more than 8% smaller than the value prescribed by said standards and safety regulations.
7. The method as claimed in claim 1, characterized in that said percentage is by at most 6% smaller than the value pre-scribed by said standards and safety regulations.
8. The method as claimed in claims 1 - 3, characterized in that said percentage is by at most about 5% smaller than the value prescribed by said standards and safety regulations.
9. A nuclear reactor comprising a control system which limits the maximum permitted effective (nominal) output power in continuous operation and an emergency cooling system having a cooling capacity proportioned to said nominal power output, characterized in that said control system is devised so that the maximum permitted effective power output of the reactor is up to at least 5% larger than that, which follows from the cooling capacity of the emergency cooling system according to the nuclear standards and safety regulations valid in 1984.
10. A nuclear reactor comprising a control system, which limits the maximum nominal effective power output in normal operation, and an emergency cooling system having a cooling capacity proportioned to said nominal power output, characterized in that said control system is designed for a maximum and effective power output in normal operation, which exceeds the maximum and effective power output permitted by the standard DIN 25436.
11. A nuclear reactor comprising a control system, which limits the maximum nominal effective power output in normal operation, and an emergency cooling system having a cooling capacity proportioned to said nominal power output, characterized in that said control system is designed for a maximum and effective power output in normal operation, which exceeds the maximum and effective power output permitted by the ANS standard ANSI/ANS-5.1-1979/1985.
12. A nuclear reactor according to claim 10 or 11, characterized in that, the maximum and effective power output is by at most 5%
larger than the value according to said standard.
larger than the value according to said standard.
13. A nuclear reactor according to claim 10 or 11, characterized by: The maximum and effective power output is by at most 6% larger than the value according to said standard.
14. A nuclear reactor according to claims 10 or 11 characterized, in that the maximum and effective power output is by at least 1%
larger than the maximum and effective power output according to said standard.
larger than the maximum and effective power output according to said standard.
15. A nuclear reactor according to claims 10 or 11, characterized in that the maximum and effective power output is by at least 2 %
larger than the maximum and effective power output according to said standard.
larger than the maximum and effective power output according to said standard.
16. A nuclear reactor according to claim 10 or 11, characterized in that the maximum and effective power output is by at least 3%
larger than the maximum and effective power output according to said standard.
larger than the maximum and effective power output according to said standard.
17. A nuclear reactor according to claim 10 or 11 characterized in that the maximum and effective power output is by at least 4%
larger than the maximum and effective power output according to said standard.
larger than the maximum and effective power output according to said standard.
18. A nuclear reactor having a given maximum and effective power output (installed nominal capacity) and comprising an emergency cooling system of a predetermined cooling capacity for removal of the decay heat, where the decay heat-cooling capacity is a given percentage of the installed nominal capacity, characterized in that said percentage is smaller than the value according to the safety regulations valid in 1984.
19. A nuclear reactor according to claim 18, characterized in that said percentage is by at most 10% smaller than the value according to said safety regulations.
20. A nuclear reactor according to claim 18, characterized in that said percentage is by at most 8% smaller than the value according to said safety regulations.
21. A nuclear reactor according to claim 18, characterized in that said percentage is by at most 6% smaller than the value according to said safety regulations.
22. A nuclear reactor according to claim 18, characterized in that said percentage is by at most about 5% smaller than the value according to said safety regulations.
23. A nuclear reactor according to claim 18, characterized in that said percentage is at least 1% smaller than the value according to said safety regulations.
24. A nuclear reactor according to claim 18, characterized in that said percentage is at least 2% smaller than the value according to said safety regulations.
25. A nuclear reactor according to claim 18, characterized in that said percentage is at least 3% smaller than the value according to said present safety regulations.
26. A nuclear reactor according to claim 18, characterized in that said percentage is at least 4% smaller than the value according to said safety regulations.
27. The method as claimed in claim 1, wherein said regulations are the ANSI/ANS-5.1-1979/1985 standard.
28. The method as claimed in claim 1, wherein said standards are the DIN 25463 / July 1982 standard.
29. The nuclear reactor as claimed in claim 18, wherein said regulations are in accordance with the ANSI/ANS-5.1-1979/1985 standard.
30. The nuclear reactor as claimed in claim 18, wherein said regulations are in accordance with the DIN 25463/July 1982 standard.
Applications Claiming Priority (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| DE3435541 | 1984-09-27 | ||
| DEP3435541.3 | 1984-09-27 | ||
| DE19853524175 DE3524175A1 (en) | 1984-09-27 | 1985-07-05 | METHOD FOR OPERATING A CORE REACTOR AND CORE REACTOR FOR IMPLEMENTING THE METHOD |
| DEP3524175.6 | 1985-07-05 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CA1267446A true CA1267446A (en) | 1990-04-03 |
Family
ID=25825167
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA000491578A Expired - Lifetime CA1267446A (en) | 1984-09-27 | 1985-09-26 | Nuclear reactor and method of operating same |
Country Status (3)
| Country | Link |
|---|---|
| EP (1) | EP0176105A3 (en) |
| CA (1) | CA1267446A (en) |
| DE (1) | DE3524175A1 (en) |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| DE2640786A1 (en) * | 1976-09-10 | 1978-03-16 | Hochtemperatur Reaktorbau Gmbh | Decay heat removal from pebble-bed nuclear reactor - uses auxiliary heat exchangers with gas flow reversed to natural convection |
| DE2942937C2 (en) * | 1979-10-24 | 1984-10-18 | Brown Boveri Reaktor GmbH, 6800 Mannheim | Device for residual heat removal and / or for emergency cooling of a water-cooled nuclear reactor plant |
| DE3212322A1 (en) * | 1982-04-02 | 1983-10-06 | Hochtemperatur Reaktorbau Gmbh | Method for controlling design basis and hypothetical accidents in a nuclear power station |
-
1985
- 1985-07-05 DE DE19853524175 patent/DE3524175A1/en not_active Withdrawn
- 1985-09-26 CA CA000491578A patent/CA1267446A/en not_active Expired - Lifetime
- 1985-09-27 EP EP85112294A patent/EP0176105A3/en not_active Ceased
Also Published As
| Publication number | Publication date |
|---|---|
| EP0176105A3 (en) | 1987-10-14 |
| DE3524175A1 (en) | 1986-04-03 |
| EP0176105A2 (en) | 1986-04-02 |
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