CN107153732B - Method for analyzing transient state of pressurized water reactor core by Pin-by-Pin - Google Patents

Method for analyzing transient state of pressurized water reactor core by Pin-by-Pin Download PDF

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CN107153732B
CN107153732B CN201710301187.2A CN201710301187A CN107153732B CN 107153732 B CN107153732 B CN 107153732B CN 201710301187 A CN201710301187 A CN 201710301187A CN 107153732 B CN107153732 B CN 107153732B
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CN107153732A (en
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曹良志
谢伟华
李云召
赵传奇
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Xian Jiaotong University
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Abstract

A method for analyzing a pressurized water reactor transient state by Pin-by-Pin comprises the following steps: 1. calculating the initial state of the reactor core as the initial point of transient analysis; 2. for the next time step, considering external disturbance information, and updating the section by using the neutron flux density of the previous time step; obtaining coefficients and source terms of an SP3 steady-state equation; 3. solving time discrete tth numerical value by using exponential function expansion block method based on spatial Pin-by-Pin numerical valuen+1Carrying out iterative solution on a steady SP3 equation of the step to obtain the zero-order and second-order neutron flux densities of the grid cells under the new time step after convergence; 4. updating the concentration of the precursor nucleus of the delayed neutrons; 5. judging whether the current time step is the last time step or not, if not, repeating 2-4; if yes, stopping calculation and outputting the transient parameters. Compared with the assembly homogenization method, errors introduced by power reconstruction do not need to be considered; compared with diffusion approximation, only the zeroth-order neutron flux density can be obtained, and more core information is reserved. While the solution time is not significantly increased.

Description

Method for analyzing transient state of pressurized water reactor core by Pin-by-Pin
Technical Field
The invention relates to the field of nuclear reactor design and reactor physical computation, in particular to a method for analyzing a pressurized water reactor transient state by Pin-by-Pin.
Background
The core is subject to a wide variety of event-related phenomena, which can be broadly classified into three categories, short-time transients, medium-time transients, and long-time transients, according to their process time constants, short-time transient durations are typically on the order of milliseconds to seconds, transient durations are primarily due to rapid changes in neutron flux density caused by artificially or accidentally induced reactivity, such as transient experiments, startup, shutdown processes, and the like.
The method for directly solving the neutron transport equation by using the deterministic theory method is time-consuming and cannot meet the real-time requirement, so the conventional method is to solve the neutron diffusion equation with the approximate angle P1, the assembly homogenization idea is generally adopted in space, the neutron flux density with homogenized assembly is firstly obtained, and then the power distribution in the assembly is obtained through power reconstruction. However, the neutron flux density and power distribution obtained in this way often have large errors from the actual real result, especially for local inhomogeneous effects. The SP3 approximation equation is a more accurate form of neutron transport equation relative to the neutron diffusion equation, and the neutron flux density and power distribution inside each cell can be obtained by solving the SP3 approximation equation spatially for Pin-by-Pin (typical grid size with cells). The nuclear core procedure EFEN (exponential function expansion block method) has been developed by NECP laboratories at the university of west ann for transportation to solve the SP3 approximation equation.
Therefore, the key point is how to solve the transient SP3 approximate equation by the Pin-by-Pin to obtain the neutron information in each grid cell, and the real condition of the interior of the reactor core in the transient operation process of the reactor is reflected and analyzed in real time.
Disclosure of Invention
In order to overcome the problems in the prior art, the invention aims to provide a method for analyzing the transient state of a pressurized water reactor by a Pin-by-Pin method, a difference method and an integration method are adopted to disperse a transient neutron transport equation into a steady-state neutron transport equation, the steady-state neutron transport equation is converted into a steady-state SP3 approximate equation, then the Pin-by-Pin iteration is used for solving the steady-state equation at each time step, the neutron flux density and the power distribution in the grid cells at each time step are obtained, and meanwhile, the calculation time still meets the actual application requirement.
In order to achieve the purpose, the invention adopts the following technical scheme to implement:
a method for analyzing a pressurized water reactor transient state by Pin-by-Pin comprises the following steps:
step 1: calculating the initial state of the reactor core as the initial point of transient analysis; obtaining the zero-order and second-order neutron flux densities in each grid cell and the corrected multi-group microscopic total cross section, the microscopic scattering cross section and the microscopic fission cross section; the method specifically comprises the following steps:
1) reading reactor core geometric parameters in an input card for geometric modeling, reading an original macroscopic total section, a macroscopic scattering section, a delayed neutron share, a decay constant, energy group information, a neutron energy spectrum, the number of neutrons generated by single fission, and reading transient time step information and disturbance information of nuclides;
2) calculating initial delayed neutron precursor nuclear concentration, a coefficient and a source term of an SP3 steady-state equation based on input parameters and reactor core geometry; the calculation formula is as follows:
Figure BDA0001284359330000031
in the formula:
g' -pre-collision neutron energy group;
t0-an initial time;
x-neutron position;
Sf(t0) -a fission source at an initial moment;
v (x) -number of neutrons produced by a single fission;
Σf,g'(x,t0) -a macroscopic fission cross-section of the group g' neutrons at x at an initial instant;
ν(x)Σf,g'(x,t0) -macroscopic cross section of group g' neutrons at x at an initial instant;
Figure BDA0001284359330000032
-the 0 th order g' group neutron flux density moment at the initial instant;
Figure BDA0001284359330000033
in the formula:
i-number of delayed neutron sets;
nd is total group number of delayed neutrons;
Ci(x,t0) -initial time at x the ith group of delayed neutron precursor nuclear concentrations;
λi-group i delayed neutron decay constants;
βi(x) -the ith group of delayed neutron contributions at x;
Figure BDA0001284359330000041
in the formula:
Sd-a delayed neutron source item;
χd,g,i(x) -the ith group of delayed neutron spectra at x;
Figure BDA0001284359330000042
in the formula:
D1-an equivalent diffusion coefficient;
Σt,g(x) -a macroscopic total cross-section of the g-th group of neutrons at x;
Σr,g(x)＝Σt,g(x)-Σs,g→g(x) Formula (5)
In the formula:
Σr,g(x) -a macroscopic removal cross-section of the g-th group of neutrons at x;
Σs,g→g(x) -a group g neutron macroscopic self-scattering cross-section at x;
Figure BDA0001284359330000043
in the formula:
Sg-an equivalent neutron source;
μ0-an initial angle;
g-neutron energy group after collision;
Σs,g'→g(x,μ0) -angle μ at x0Is scattered from the g 'th group into a macroscopic cross section of the g' th group;
keff-an effective multiplication factor;
χg(x) -group g neutron spectrum;
3) performing Pin-by-Pin steady-state calculation by using an exponential function expansion block method to obtain zero-order and second-order neutron flux densities and effective value-added coefficients inside each grid cell at an initial time step; the SP3 steady state equation is shown below:
Figure BDA0001284359330000051
in the formula:
Figure BDA0001284359330000052
-the 2 nd order g' group neutron flux density moment at the initial instant;
4) and correcting the original section by using the effective incremental coefficient to obtain updated section parameters, wherein the updated section parameters are shown as a formula:
Figure BDA0001284359330000053
step 2: for the next time step, considering external disturbance information, and updating the section by using the neutron flux density of the previous time step; obtaining coefficients and source terms of an SP3 steady-state equation;
Figure BDA0001284359330000054
in the formula:
Figure BDA0001284359330000055
-a macroscopic total cross section of equivalent neutrons at x;
Figure BDA0001284359330000056
in the formula:
Figure BDA0001284359330000057
-group g equivalent macroscopic removal cross-section at x;
Vg(x) -group g neutron velocities at x;
Δtn＝tn+1-tn-step size from nth to n +1 time step;
Figure BDA0001284359330000061
Figure BDA0001284359330000062
——tn(ii) the 0 th order g' group neutron flux density moment at time;
Figure BDA0001284359330000063
-an equivalent group g neutron spectrum at x;
Sf(tn+1) -fission source at x at the n +1 th time step;
Ci(x,tn)——tnthe ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Figure BDA0001284359330000064
-the coefficients of the fission source at the x position at the (n + 1) th time step;
Figure BDA0001284359330000065
in the formula:
γ -ratio of nth time step to nth-1 time step;
Δtn-1＝tn-tn-1-step sizes from the n-1 th to the n-th time step;
κ0(λi) -a custom function;
κ1(λi) -a custom function;
κ2(λi) -a custom function;
and step 3: solving time discrete tth numerical value by using exponential function expansion block method based on spatial Pin-by-Pin numerical valuen+1Carrying out iterative solution on a steady SP3 equation of the step to obtain the zero-order and second-order neutron flux densities of the grid cells under the new time step after convergence; the steady state SP3 equation is as follows:
Figure BDA0001284359330000071
in the formula:
Figure BDA0001284359330000072
——tn+1(ii) the 0 th order g' group neutron flux density moment at time;
Figure BDA0001284359330000073
——tn+1(ii) the 2 nd order g' group neutron flux density moment at time;
Figure BDA0001284359330000074
——tn(ii) the 2 nd order g' group neutron flux density moment at time;
the result error of the effective multiplication coefficient and the flux of the reactor core at two adjacent time steps is less than
The user considers convergence when giving a convergence limit; the convergence criterion formula is as follows:
Figure BDA0001284359330000075
in the formula:
error-effective incremental coefficient error;
keff(tn+1) At tn+1Calculating the effective value-added coefficient at every moment;
keff(tn) At tnCalculating the effective value-added coefficient at every moment;
errorf 0-0 order neutron flux density moment error;
Figure BDA0001284359330000076
at tn+1Calculating 0-order neutron flux density moment at any moment;
Figure BDA0001284359330000077
at tnCalculating 0-order neutron flux density moment at any moment;
errorf 2-2 nd order neutron flux density moment error;
Figure BDA0001284359330000078
at tn+1Calculating 2-order neutron flux density moment at any moment;
Figure BDA0001284359330000079
at tnCalculating 2-order neutron flux density moment at any moment;
and 4, step 4: updating the concentration of the delayed neutron precursor nucleus, wherein the formula is as follows:
formula (15)
In the formula:
Ci(x,tn+1)——tn+1the ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Sf(tn) -fission source at x at nth time step;
Sf(tn-1) -fission source at x at nth time step;
and 5: judging whether the current time step is the last time step or not, if not, repeating 2-4; if yes, stopping calculation and outputting transient parameters; and timely and accurate reactor operation transient information is provided for timely operation of a reactor operator.
Compared with the prior art, the invention has the following advantages:
1. and (3) solving by adopting Pin-by-Pin in space, the neutron flux density and power distribution of each grid cell can be obtained, and errors introduced by power reconstruction do not need to be considered relative to an assembly homogenization method.
2. The SP3 approximation is adopted in the angle, the zero-order and second-order neutron flux densities can be obtained, only the zero-order neutron flux density can be obtained compared with the diffusion approximation, and more core information is reserved. While the solution time is not significantly increased.
Drawings
FIG. 1 is a flow chart of the operation of a Pin-by-Pin analysis for PWR transients.
Detailed Description
The invention is described in further detail below with reference to the following figures and detailed description:
the invention adopts a difference method and an integration method to disperse a transient neutron transport equation into a steady-state neutron transport equation, then converts the steady-state neutron transport equation into a steady-state SP3 approximate equation, and then solves the steady-state equation in a Pin-by-Pin iteration mode at each time step to obtain the neutron flux density and power distribution in the grid cells at each time step, and meanwhile, the calculation time still meets the actual application requirement.
In order to achieve the purpose, the invention adopts the following technical scheme to implement:
a method for analyzing a pressurized water reactor transient state by Pin-by-Pin comprises the following steps:
step 1: calculating the initial state of the reactor core as the initial point of transient analysis; obtaining the zero-order and second-order neutron flux densities in each grid cell and the corrected multi-group microscopic total cross section, the microscopic scattering cross section and the microscopic fission cross section; the method specifically comprises the following steps:
1) reading reactor core geometric parameters in an input card for geometric modeling, reading an original macroscopic total section, a macroscopic scattering section, a delayed neutron share, a decay constant, energy group information, a neutron energy spectrum, the number of neutrons generated by single fission, and reading transient time step information and disturbance information of nuclides;
2) calculating initial delayed neutron precursor nuclear concentration, a coefficient and a source term of an SP3 steady-state equation based on input parameters and reactor core geometry; the calculation formula is as follows:
Figure BDA0001284359330000091
in the formula:
g' -pre-collision neutron energy group;
t0-an initial time;
x-neutron position;
Sf(t0) -a fission source at an initial moment;
v (x) -number of neutrons produced by a single fission;
Σf,g'(x,t0) -a macroscopic fission cross-section of the group g' neutrons at x at an initial instant;
ν(x)Σf,g'(x,t0) -macroscopic cross section of group g' neutrons at x at an initial instant;
Figure BDA0001284359330000092
-the 0 th order g' group neutron flux density moment at the initial instant;
Figure BDA0001284359330000093
in the formula:
i-number of delayed neutron sets;
nd is total group number of delayed neutrons;
Ci(x,t0) -initial time at x the ith group of delayed neutron precursor nuclear concentrations;
λi-group i delayed neutron decay constants;
βi(x) -the ith group of delayed neutron contributions at x;
Figure BDA0001284359330000101
in the formula:
Sd-a delayed neutron source item;
χd,g,i(x) -the ith group of delayed neutron spectra at x;
Figure BDA0001284359330000102
in the formula:
D1-an equivalent diffusion coefficient;
Σt,g(x) -a macroscopic total cross-section of the g-th group of neutrons at x;
Σr,g(x)＝Σt,g(x)-Σs,g→g(x) Formula (5)
In the formula:
Σr,g(x) -a macroscopic removal cross-section of the g-th group of neutrons at x;
Σs,g→g(x) -a group g neutron macroscopic self-scattering cross-section at x;
Figure BDA0001284359330000103
in the formula:
Sg-an equivalent neutron source;
μ0-an initial angle;
g-neutron energy group after collision;
Σs,g'→g(x,μ0) -angle μ at x0Is scattered from the g 'th group into a macroscopic cross section of the g' th group;
keff-an effective multiplication factor;
χg(x) -group g neutron spectrum;
3) performing Pin-by-Pin steady-state calculation by using an EFEN program (exponential function expansion block method) to obtain zero-order and second-order neutron flux densities and effective value-added coefficients inside each grid cell at an initial time step; the SP3 steady state equation is shown below:
Figure BDA0001284359330000111
in the formula:
Figure BDA0001284359330000112
-the 2 nd order g' group neutron flux density moment at the initial instant;
4) correcting the original section by using the effective incremental coefficient to obtain updated section parameters, such as formula
Shown in the figure:
Figure BDA0001284359330000113
step 2: for the next time step, considering external disturbance information, and updating the section by using the neutron flux density of the previous time step; obtaining coefficients and source terms of an SP3 steady-state equation;
Figure BDA0001284359330000114
in the formula:
Figure BDA0001284359330000115
-a macroscopic total cross section of equivalent neutrons at x;
Figure BDA0001284359330000116
in the formula:
Figure BDA0001284359330000117
-group g equivalent macroscopic removal cross-section at x;
Vg(x) -group g neutron velocities at x;
Δtn＝tn+1-tn-step size from nth to n +1 time step;
Figure BDA0001284359330000121
Figure BDA0001284359330000122
——tn(ii) the 0 th order g' group neutron flux density moment at time;
Figure BDA0001284359330000123
-an equivalent group g neutron spectrum at x;
Sf(tn+1) -fission source at x at the n +1 th time step;
Ci(x,tn)——tnthe ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Figure BDA0001284359330000124
-the coefficients of the fission source at the x position at the (n + 1) th time step;
Figure BDA0001284359330000125
in the formula:
γ -ratio of nth time step to nth-1 time step;
Δtn-1＝tn-tn-1-step sizes from the n-1 th to the n-th time step;
κ0(λi) -a custom function;
κ1(λi) -a custom function;
κ2(λi) -a custom function;
and step 3: solving the t-th value of time dispersion by using the numerical value of Pin-by-Pin in the space of an exponential function expansion block method (EFEN program)n+1Carrying out iterative solution on a steady SP3 equation of the step to obtain the zero-order and second-order neutron flux densities of the grid cells under the new time step after convergence; the steady state SP3 equation is as follows:
Figure BDA0001284359330000131
in the formula:
Figure BDA0001284359330000132
——tn+1(ii) the 0 th order g' group neutron flux density moment at time;
Figure BDA0001284359330000133
——tn+1(ii) the 2 nd order g' group neutron flux density moment at time;
Figure BDA0001284359330000134
——tn(ii) the 2 nd order g' group neutron flux density moment at time;
and (4) judging that the effective multiplication coefficient and the flux of the reactor core are converged when the result error of the effective multiplication coefficient and the flux of the reactor core in two adjacent time steps is less than the convergence limit given by the user. The convergence criterion formula is as follows:
Figure BDA0001284359330000135
in the formula:
error-effective incremental coefficient error;
keff(tn+1) At tn+1Calculating the effective value-added coefficient at every moment;
keff(tn) At tnCalculating the effective value-added coefficient at every moment;
errorf 0-0 order neutron flux density moment error;
Figure BDA0001284359330000136
at tn+1Calculating 0-order neutron flux density moment at any moment;
Figure BDA0001284359330000137
at tnCalculating 0-order neutron flux density moment at any moment;
errorf 2-2 nd order neutron flux density moment error;
Figure BDA0001284359330000138
at tn+1Calculating 2-order neutron flux density moment at any moment;
Figure BDA0001284359330000141
at tnCalculating 2-order neutron flux density moment at any moment;
and 4, step 4: updating the concentration of the delayed neutron precursor nucleus, wherein the formula is as follows:
Figure BDA0001284359330000142
in the formula:
Ci(x,tn+1)——tn+1the ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Sf(tn) -fission source at x at nth time step;
Sf(tn-1) -fission source at x at nth time step;
and 5: judging whether the current time step is the last time step or not, if not, repeating 2-4; if yes, stopping calculation and outputting transient parameters; and timely and accurate reactor operation transient information is provided for timely operation of a reactor operator.

Claims (1)

1. A method for analyzing a pressurized water reactor transient state by a Pin-by-Pin is characterized by comprising the following steps: the method comprises the following steps:
step 1: calculating the initial state of the reactor core as the initial point of transient analysis; obtaining the zero-order and second-order neutron flux densities in each grid cell and the corrected multi-group microscopic total cross section, the microscopic scattering cross section and the microscopic fission cross section; the method specifically comprises the following steps:
1) reading reactor core geometric parameters in an input card for geometric modeling, reading an original macroscopic total section, a macroscopic scattering section, a delayed neutron share, a decay constant, energy group information, a neutron energy spectrum, the number of neutrons generated by single fission, and reading transient time step information and disturbance information of nuclides;
2) calculating initial delayed neutron precursor nuclear concentration, a coefficient and a source term of an SP3 steady-state equation based on input parameters and reactor core geometry; the calculation formula is as follows:
Figure FDA0002396590250000011
in the formula:
g' -pre-collision neutron energy group;
t0-an initial time;
x-neutron position;
Sf(t0) -a fission source at an initial moment;
v (x) -number of neutrons produced by a single fission;
Σf,g'(x,t0) -a macroscopic fission cross-section of the group g' neutrons at x at an initial instant;
ν(x)Σf,g'(x,t0) -macroscopic cross section of group g' neutrons at x at an initial instant;
Figure FDA0002396590250000012
-the g' th group of neutrons of order 0 at the initial momentMoment of mass density;
Figure FDA0002396590250000013
in the formula:
i-number of delayed neutron sets;
nd is total group number of delayed neutrons;
Ci(x,t0) -initial time at x the ith group of delayed neutron precursor nuclear concentrations;
λi-group i delayed neutron decay constants;
βi(x) -the ith group of delayed neutron contributions at x;
Figure FDA0002396590250000021
in the formula:
Sd-a delayed neutron source item;
χd,g,i(x) -the ith group of delayed neutron spectra at x;
Figure FDA0002396590250000022
in the formula:
D1-an equivalent diffusion coefficient;
Σt,g(x) -a macroscopic total cross-section of the g-th group of neutrons at x;
Σr,g(x)＝Σt,g(x)-Σs,g→g(x) Formula (5)
In the formula:
Σr,g(x) -a macroscopic removal cross-section of the g-th group of neutrons at x;
Σs,g→g(x) -a group g neutron macroscopic self-scattering cross-section at x;
Figure FDA0002396590250000023
in the formula:
Sg-an equivalent neutron source;
μ0-an initial angle;
g-neutron energy group after collision;
Σs,g'→g(x,μ0) -angle μ at x0Is scattered from the g 'th group into a macroscopic cross section of the g' th group;
keff-an effective multiplication factor;
χg(x) -group g neutron spectrum;
3) performing Pin-by-Pin steady-state calculation by using an exponential function expansion block method to obtain zero-order and second-order neutron flux densities and effective value-added coefficients inside each grid cell at an initial time step; the SP3 steady state equation is shown below:
Figure FDA0002396590250000031
in the formula:
Figure FDA0002396590250000032
-the 2 nd order g' group neutron flux density moment at the initial instant;
4) and correcting the original section by using the effective incremental coefficient to obtain updated section parameters, wherein the updated section parameters are shown as a formula:
Figure FDA0002396590250000033
step 2: for the next time step, considering external disturbance information, and updating the section by using the neutron flux density of the previous time step; obtaining coefficients and source terms of an SP3 steady-state equation;
Figure FDA0002396590250000034
in the formula:
Figure FDA0002396590250000035
-a macroscopic total cross section of equivalent neutrons at x;
Figure FDA0002396590250000036
in the formula:
Figure FDA0002396590250000037
-group g equivalent macroscopic removal cross-section at x;
Vg(x) -group g neutron velocities at x;
Δtn＝tn+1-tn-step size from nth to n +1 time step;
Figure FDA0002396590250000041
Figure FDA0002396590250000042
——tn(ii) the 0 th order g' group neutron flux density moment at time;
Figure FDA0002396590250000043
-an equivalent group g neutron spectrum at x;
Sf(tn+1) -fission source at x at the n +1 th time step;
Ci(x,tn)——tnthe ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Figure FDA0002396590250000044
-the coefficients of the fission source at the x position at the (n + 1) th time step;
Figure FDA0002396590250000045
in the formula:
γ -ratio of nth time step to nth-1 time step;
Δtn-1＝tn-tn-1-step sizes from the n-1 th to the n-th time step;
κ0(λi) -a custom function;
κ1(λi) -a custom function;
κ2(λi) -a custom function;
and step 3: solving time discrete tth numerical value by using exponential function expansion block method based on spatial Pin-by-Pin numerical valuen+1Carrying out iterative solution on a steady SP3 equation of the step to obtain the zero-order and second-order neutron flux densities of the grid cells under the new time step after convergence; the steady state SP3 equation is as follows:
Figure FDA0002396590250000051
in the formula:
Figure FDA0002396590250000052
——tn+1(ii) the 0 th order g' group neutron flux density moment at time;
Figure FDA0002396590250000053
——tn+1(ii) the 2 nd order g' group neutron flux density moment at time;
Figure FDA0002396590250000054
——tn(ii) the 2 nd order g' group neutron flux density moment at time;
the result errors of the effective multiplication coefficient and the flux of the reactor core in two adjacent time steps are smaller than the convergence limit given by the user, and then the reactor core is considered to be converged; the convergence criterion formula is as follows:
Figure FDA0002396590250000055
in the formula:
error-effective incremental coefficient error;
keff(tn+1) At tn+1Calculating the effective value-added coefficient at every moment;
keff(tn) At tnCalculating the effective value-added coefficient at every moment;
errorf 0-0 order neutron flux density moment error;
Figure FDA0002396590250000056
at tn+1Calculating 0-order neutron flux density moment at any moment;
Figure FDA0002396590250000057
at tnCalculating 0-order neutron flux density moment at any moment;
errorf 2-2 nd order neutron flux density moment error;
Figure FDA0002396590250000058
at tn+1Calculating 2-order neutron flux density moment at any moment;
Figure FDA0002396590250000059
at tnCalculating 2-order neutron flux density moment at any moment;
and 4, step 4: updating the concentration of the delayed neutron precursor nucleus, wherein the formula is as follows:
Figure FDA0002396590250000061
in the formula:
Ci(x,tn+1)——tn+1the ith group of delayed neutron precursor nucleus concentration at the x position at the moment;
Sf(tn) -fission source at x at nth time step;
Sf(tn-1) -the fission source at time step n-1 at x;
and 5: judging whether the current time step is the last time step or not, if not, repeating 2-4; if yes, stopping calculation and outputting transient parameters; and timely and accurate reactor operation transient information is provided for timely operation of a reactor operator.
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