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Signal processing system design for improved shutdown system 1
of CANDU reactor in large break of LOCA events 2
3
Hossam A. Gabera,b*, Lingzhi Xiaa, Manir U. Ishama, Vladimir Ponomarevc 4
5 a Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of 6
Technology, 2000 Simcoe Street North, Oshawa, L1H 7K4 ON, Canada 7 b Faculty of Engineering and Applied Science, University of Ontario Institute of 8
Technology, 2000 Simcoe Street North, Oshawa, L1H 7K4 ON, Canada 9 c Megawatt Solutions Inc., Ontario, Canada 10
11
HIGHLIGHTS 12
An innovative scheme of the neutronic signal processing system to improve the CANDU 13
reactor SDS1 performance is proposed. 14
A point kinetic model is developed to fundamentally analyze the CANDU large break of 15
LOCA transient. 16
A MATLAB/Simulink simulation platform is established to model the existing CANDU 17
SDS1 signal processing system and its trip logics. 18
A specific design of the proposed signal processing system for CANDU SDS1 is performed 19
and implemented. 20
21
ABSTRACT 22
For CANDU reactors, several options to improve CANDU nuclear power plant operation 23
safety margin have been investigated in this paper. A particular attention is paid to the 24
response time of CANDU shutdown system number 1 (SDS1) in case of large break loss 25
of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental 26
analysis is performed to CANDU LLOCA event, and the power transient signal is 27
generated. In order to improve the SDS1 response time during LLOCA events, an 28
innovative power measurement and signal processing system is particularly designed. The 29
new signal processing system is implemented with the input of the LLOCA power transient, 30
and the simulation results of the reactor trip time and signal are compared to those of the 31
existing system in CANDU power plants. It is demonstrated that the new signal processing 32
system can not only achieve a shorter reactor trip time than the existing system, but also 33
accommodate the spurious trip immunity. This will significantly enhance the safety margin 34
for the power plant operation, or bring extra economical benefits to the power plant units. 35
2
36
Keywords: CANDU, SDS1, LOCA, power measurement, signal processing, safety margin, 37
reactor trip 38
39
40
1. Introduction 41
1.1 CANDU reactor safety systems 42
In the world, the nuclear power plant must be equipped with safety systems, such that if 43
any anticipated or unanticipated accident happens, the plant operation is maintained 44
without affecting the public safety. Major functions of the nuclear safety systems are to 45
prevent the plant physical barriers from damage and prevent the radioactive substances 46
from releasing. Furthermore, the safety systems have to mitigate the hazard caused by 47
potential accidents in order to reduce or maintain the post-accident impact to a level as low 48
as possible. 49
Three main objectives of nuclear reactor safety systems are to shutdown the reactor, 50
maintain it in a cooled condition, and prevent the release of radioactive material. For 51
CANDU reactors, the major safety systems are categorized into reactor shutdown systems, 52
emergency core cooling system, and containment systems. Shutdown systems are used to 53
stop the chain reactor and shut down the reactor operation. Emergency core cooling system 54
is used to refill the reactor fuel channels with coolant and remove residual or decay heat 55
from the fuel. Containment systems could prevent radioactivity release, which may escape 56
from the reactor core, to the environment. 57
CANDU reactor designs have a typical positive coolant void coefficient, as well as a small 58
power coefficient. This implies that steam generated in the reactor coolant will increase the 59
reaction rate, which in turn will increase the reactor power and generate more steam. This 60
is usually considered as a negative feature in the power reactor design. The reactivity 61
devices, such as adjustor rods, liquid zone controllers, and mechanical control absorbers 62
are capable to fulfill the responsibility of compensating the positive or negative reactivity 63
so as to control the reactor power in an appropriate rate or maintain the reactor operated 64
within an acceptable power range. However, when an accident happens, the induced 65
reactivity represents a fast change rate and a relatively large change range, which 66
significantly exceeds the regulating ability of those reactivity devices. An extreme scenario 67
is the large break of loss of coolant accident (LLOCA), which induces very fast positive 68
reactivity insertion such that the reactivity devices cannot meet the requirement to quickly 69
resist the reactor power increase [1]. Therefore, the shutdown systems are designed to meet 70
the speed requirements of such a scenario. In CANDU reactors, there are two separate 71
shutdown systems: shutdown system number 1 (SDS1), and shutdown system number 2 72
3
(SDS2). Each of them performs the function of fast shutting down the reactor under 73
LLOCA conditions independently, and both of them are triggered by the 2-out-of-3 (2oo3) 74
redundancy logic from two separate detection systems, respectively. Fig. 1 represents the 75
SDS1 and SDS2 located in CANDU reactors [2]. Reactor operation is terminated when a 76
neutronic or process parameter reaches an unacceptable range. Design of two shutdown 77
systems in CANDU reactors represents that postulated accident events coincident with 78
failure of shutdown are incredible, and consequently beyond the design basis. 79
80
81
Fig. 1 CANDU SDS1 and SDS2 [2] 82
83
SDS1 terminates the reactor operation and maintains the reactor in a safe condition by 84
inserting 28 spring-assisted shutoff rods from the top of the reactor calandria to the reactor 85
core. The system has sufficient speed and negative reactivity depth to reduce the reactor 86
power to levels consistent with available cooling. SDS2 rapidly injects its high-pressure 87
liquid poison, a strong neutron absorbing solution to perform the shutdown action. The 88
parameters chosen are different from those used for SDS1. 89
4
1.2 Research motivation 90
As one of the safety systems in CANDU nuclear power plant, SDS1 has to meet strict 91
requirements: shutdown the reactor and keep it subcritical whenever necessary; have high 92
availability; have online testing ability; have enough redundancy and independency; 93
perform its function on time whenever necessary [3]. 94
To obtain the mentioned qualifications, SDS1 is designed to be a triplicate, relay logic 95
applied system [4]. There are a total of three trip channels (D, E, and F) with completely 96
independent and physically separated power supplies, trip parameter sensors, 97
instrumentation trip logic and annunciation. Each trip channel has exactly the same 98
functionality. The reliability and availability criteria are met with the triple redundancy 99
while the online testing ability is allowed by the independence between each channel. 100
Meanwhile spurious trips are also effectively prevented through a two out of three (2oo3) 101
vote of the three outputs of the triple redundant trip channels. This majority voting logic 102
permits the reactor trip signal to be released only when at least two trip channels are on trip 103
status, which reduces the probability of a false trip decision. 104
According to the criteria designed by CNSC, the unavailability of CANDU SDS1 is 105
required to be less than 10-3 years per year [3]. The online testing ability is required to 106
ensure the availability of SDS1 such that the testing of SDS1 can be carried out without a 107
reduction in the effectiveness of the system. Sufficient redundancy and independency allow 108
the SDS1 to remain functional when a failure of any single component in the SDS1 happens. 109
On-time actuation of SDS1 is critical to plant safety since the consequence could be much 110
worse with a delayed shutdown in an accident with rapid transient. The response time of 111
SDS1 is the key factor that affects the shutdown speed. The shorter the response time is, 112
the faster the SDS1 can shutdown the reactor, resulting in a lower power surge. Thus 113
shortening the SDS1 response time could help improve the safety margin of the reactor 114
operation. On the other hand, if the safety margin remains unchanged, shortening the SDS1 115
response time could help increase the reactor operation power, which brings more 116
economic benefits for power plants. 117
With respect to how to improve the CANDU reactor safety margin, or the response time of 118
SDS1, some works have been done over its history. A CANDU Owners Group Inc. (COG) 119
report has specified the large LOCA safety margins in CANDU reactors with respect to the 120
challenging of it inherent positive void effect [1]. The report particularly describes the 121
continuing efforts and the raised solutions which have been made to potentially improve 122
the safety margin of reactor operation. With respect to the originally conservative CANDU 123
reactor designs, some theoretical methodologies such as Best-Estimate and Analysis of 124
Uncertainty Methodology (BEAU) and new break opening modeling are proposed. 125
Moreover, potential improvements in safety margins through physics design change are 126
discussed, such as reducing the peak reactivity during the first few seconds of a LLOCA 127
5
either through reducing the positive coolant void reactivity or increasing the rate of 128
negative reactivity addition. 129
In the recent researches, more sight is focused on promoting the SDS1 response time. The 130
response time of SDS1 is composed of sensor response time, trip logic decision-making 131
time, trip relay logic time, and the time needed to fully insert the shutoff rods into the core. 132
Primarily, the FPGA hardware technique is used to reduce the trip logic decision-making 133
time for CANDU reactors. An FPGA-based shutdown system for CANDU6 reactors is 134
designed and implemented in a hardware-in-the-loop (HIL) environment by connecting it 135
to a nuclear power plant simulator [5]. The test results are compared against those of a 136
software-based PLC implementation of the same trip logic. It is shown that the FPGA 137
implementation can shorten the response time of software-based SDS implementation by 138
as much as 86.66%. In a follow-up work [6], a CATHENA [7] thermal-hydraulic reactor 139
model is established and used to simulations, which illustrates the functional relationship 140
between the power peak values during a trip transient and the response time of the 141
shutdown systems. Then the potential benefit of improving the safety margin of the reactor 142
operation is quantified through a FPGA-based shutdown system implementation, which is 143
validated by a HIL simulation environment. A review of the current state of FPGA Systems 144
in nuclear instrumentation and control is provided in [8]. The research on FPGA 145
applications in shutdown system and online monitoring is reviewed not only for CANDU 146
reactors, but also for PWRs and BWRs. 147
Besides this, another important area, in which the potential benefit will be obtained, is to 148
look for fasting the sensor response time, i.e. improving the in-core flux detector 149
measurements for power transients. B. Arsenault specified the CANDU6 in-core detector 150
systems in [9]. A comprehensive test of the dynamics of the shutdown systems including 151
their flux detectors is performed and reported in [10]. The time required for the signals to 152
drop to a predesigned trip fraction is plotted as a function of the detectors’ position. This 153
provides a good reference for researchers to potentially improve the shutdown system by 154
optimizing the detectors’ position. For PWR reactors, a new type of inconel self-powered 155
neutron detector is designed to replace the current Pt in-core detectors [11]. This inconel 156
detector shows good signal-to-noise ratio, which also helps to increase the detector’s 157
response time so as to improve the shutdown system performance. Several filtering 158
methods are used to compensate the delayed signals of Rhodium self-powered neutron 159
detectors in PWR [12-14]. Based on these works, three digital dynamic compensation 160
methods using robust filtering are investigated and manipulated with respect to improving 161
both Rhodium and Vanadium self-powered neutron detectors’ performance, such that they 162
can provide an apparent response speed with the noise suppressed [15-16]. This helps to 163
perform an accurate on-line monitoring function for reactor core surveillance. However, 164
for reactor shutdown process with a very fast response, these methods cannot be 165
significantly taken into account. 166
6
1.3 Objectives and approaches 167
SDS1 is one of the most important safety systems in CANDU nuclear power plants since 168
it provides an effective and reversible shutdown process. Due to its importance to the plant 169
safety, the design basis events for SDS1 are designated as: loss of regulation (LOR), loss 170
of coolant accidents (LOCA), loss of coolant flow (loss of Class IV power), loss of 171
secondary side heat sinks, and loss of moderator cooling [17]. Within these design basis 172
events, LOCA represents the most severe status of the postulated accident within the core. 173
Among LOCA accidents, LLOCA illustrates the extreme conditions. As it is mentioned 174
above, shortening the SDS1 response time could help improve the safety features in the 175
plant. Therefore, with a purpose of enhancing plant safety, the current research work 176
focuses on how to improve SDS1 fast response performance during LLOCA using software 177
simulation and digital hardware implementation technology. 178
An innovative idea to improve the reactor power measurement sensitivity during large 179
LOCA accident transients is proposed [18]. The option is to develop a new faster neutronic 180
signal processing system using the current in-core neutron detectors’ log rate readings. 181
Implementation of this system requires additional hardware to the existing SDS 182
instrumentation; however it still uses the existing in-core neutron flux detectors and ion 183
chambers. Implementation of this option will prove to be very useful as it is an economical 184
solution for already operating CANDU reactors, especially for those units that are more 185
susceptible to reaching prompt criticality in case of LLOCA accidents due to their specific 186
design features of the Heat Transport System (HTS). 187
The design of the faster neutronic signal detection system will be detailed. The procedure 188
is to build a software simulation platform which is used to demonstrate the accessibility 189
and reliability of the proposed idea. MATLAB/Simulink software simulation environment 190
is employed. Implementation and commissioning will be performed. The simulation results 191
will be compared with the results introduced by the existing system. This will demonstrate 192
the advantaged performance of the new signal processing systems. 193
The structure of this paper is as follows. Section 1 introduces the CANDU safety status and 194
the motivation of this research, as well as the proposed research scope and methodology. 195
Section 2 focuses on the numerical simulation and analysis of the CANDU reactor power 196
transients during the LLOCA event. Point kinetic method is employed. Section 3 illustrates 197
the function of the existing signal processing system. MATLAB/Simulink simulation 198
platform is established to generate the trip signal when LLOCA happens. Section 4 199
represents the specific design of the new signal processing system and its implementations. 200
Simulation result of the reactor trip time is compared to that of the existing system. 201
Furthermore, potential benefits resulted in by the new system are evaluated. Conclusion 202
and discussions are represented in Section 5. 203
7
2. Power transient studies during the large break of LOCA (LLOCA) 204
event 205
2.1 Interpretation of the LLOCA event 206
A typical LLOCA event of CANDU reactors is referenced in [1]. The case particularly 207
describes the normalized peak bundle power following a 100% break in a coolant pump 208
suction pipe. Accident description starts with some assumed unfavorable conditions as 209
required by licensing, and at the same time the thermal-hydraulics analytical models are 210
chosen to be conservative to produce the maximum coolant void reactivity. 211
Within 2 seconds following a large break steam is produced in the reactor core and coolant 212
is ejected from both reactor inlet and outlet headers. Trip signals including high neutron 213
flux and high log rates are issued in about 400 milliseconds. Peak reactivity is +4.3 mk at 214
900 milliseconds. Following this event, the reactor shut-down rods begin to enter the core 215
at 900 milliseconds. It takes the shut-down rods 2 seconds to drop into the reactor core till 216
it is fully inserted to perform the maximum negative reactivity compensation, -80 mk. It is 217
mentioned that, shutdown system decreases the reactor reactivity to -69 mk at 3.16 seconds. 218
In other words, the net positive reactivity (internal feedback) of the core at this time can be 219
evaluated. Furthermore, it is also represented that, the total positive reactivity addition at 220
the time of 10 seconds is about +15mk, which should be removed by the shutdown rods. 221
Thus, the reactor is rendered safe. During this transient process, a peak power of 3.5 full 222
power unit (FPU) arrives at 1.16 seconds. 223
2.2 Point kinetic reactor modeling for the LLOCA transient 224
There are several methods to deal with modeling and simulation for CANDU reactor 225
LLOCA transient. Usually, Canadian industry units, such as AECL or Candu Energy Inc. 226
employ the industry standard toolset (IST) software, such as CATHENA to perform the 227
deterministic safety analysis [7]. The analysis involves the modeling and simulation for 228
reactor components, loop dynamics, postulated accident events, and the invoked 229
mathematical or logic algorithms during the transients. The original idea to create 230
CATHENA is producing a modeling and simulation software system as generically as 231
possible for wide range of applications, which includes either the thermal-hydraulic 232
simulation or the deterministic safety analysis. Furthermore, its oriented objective is not 233
purely limited to CANDU reactors. Therefore, the code system defines different systematic 234
and functional modules, such that it is easy for users to build the models by inputting the 235
formatted records as per the requirements. However, for the industry commercial analysis, 236
it requires about tens of thousands records to detail the nuclear power plant systems 237
including reactor dynamic model, reactor regulating system, primary heat transport system, 238
8
main steam supply system, and so on. Apparently, this is neither affordable, nor necessary 239
for the application in this research, since the investigated target is mainly focused on the 240
reactor shutdown system. Therefore, the problem should be simplified to simulate the 241
essential modules. 242
The objective of this research is to evaluate and improve the power measurement and its 243
related signal processing sensitivity which is committed to shorten the reactor trip time 244
during LLOCA event. It is suggested that a typical LLOCA transient is simulated and the 245
power transient curve can be referenced for the following analysis works. The reasonable 246
and accessible method is to employ a point kinetic reactor model equipped with the 247
approximated internal and external reactivity inputs. By this means, it prevents the 248
simulation from the utilizations of large commercial software such as CATHENA; while 249
the alternative option such as MATLAB/Simulink software simulation could be adopted, 250
which additionally provides more generic profile to researchers in other areas. 251
The point kinetic reactor model in this research refers to the CATHENA theory manual 252
[19]. Considering single energy (thermal) group and six groups of delayed neutron 253
precursors, the model equations can be represented by 254
𝑑𝑛
𝑑𝑡=𝑘(1−𝛽)−1
𝜏𝑛 + ∑ 𝜆𝑖𝐶𝑖
6𝑖=1 (1) 255
𝑑𝐶𝑖
𝑑𝑡=𝛽𝑖𝑘
𝜏𝑛 − 𝜆𝑖𝐶𝑖 (2) 256
where n is the normalized reactor power at time t; Ci is the ith group delayed neutron 257
concentration; βi is the ith group delayed neutron fraction; λi is the ith group delayed neutron 258
decay constant; β=Σβi for i= 1, 2, 3, . . . ,6; k = (1-ρ×10−3)-1, where ρ is the total reactivity; 259
τ is the mean prompt neutron lifetime, 0.000902 second [19]. 260
Since LLOCA represents very fast dynamic response, the Xenon effect is ignored in this 261
simulation. This is reasonable, because usually the time magnitude of Xenon effect 262
accumulation is much longer than the LLOCA transient range, several seconds. 263
2.3 MATLAB/Simulink modeling and simulation for the LLOCA 264
transient 265
In order to borrow the MATLAB’s advanced features such as fast and large amount of 266
matrix manipulation and calculation, the reactor kinetic model equations (1) and (2) can be 267
vectorized to 268
𝑑𝑛
𝑑𝑡=(
1
1−𝜌×10−3)(1−𝛽)−1
𝜏𝑛 + (𝜆1 … 𝜆6) (
𝐶1⋮𝐶6
) (3) 269
9
𝑑
𝑑𝑡(𝐶1⋮𝐶6
) =(
1
1−𝜌×10−3)
𝜏𝑛(𝛽1⋮𝛽6
) − (𝜆1
⋱𝜆6
)(𝐶1⋮𝐶6
) (4) 270
As far as the initial condition is considered, the initial reactor power is 1.0 FPU. The 271
accident starts from the referenced reactor operation condition, i.e., the equilibrium core 272
under the full power operation. Thus, 273
𝑛0 = 1.0 (5) 274
where 𝑛0 is the initial value of the reactor power. 275
In this way, the initial condition of delayed neutron precursors concentration can be 276
calculated by defining the left side of equation (4) to be zero, such that 277
(𝐶1⋮𝐶6
)
0
=
(
𝛽1
𝜆1𝜏
⋮𝛽6
𝜆6𝜏)
(6) 278
The neutronic parameters are represented in Table 1. 279
280
Table 1 Point reactor neutronics data for CANDU equilibrium fuel [19] 281
282
Group Delayed Neutron Fractions (βi) Decay Constants, λi(s-1)
1 0.000295 0.000612
2 0.001165 0.03155
3 0.001033 0.1218
4 0.00235 0.3175
5 0.00078 1.389
6 0.000197 3.784
283
The reactivity, ρ, contains two parts: one is the internal reactivity feedback induced by the 284
varying of reactor variables such as Xenon concentration, fuel temperature, coolant 285
temperature, and coolant density (void fraction); the other is the external dynamic reactivity 286
which is introduced by the reactivity devices such as liquid zone controllers, adjuster rods, 287
and mechanical absorbers, or the shutdown systems such as SDS1 and SDS2. For LLOCA, 288
the most concerned reactivity parts are the internal reactivity feedback by coolant density, 289
and the external dynamic reactivity caused by shutdown rods. In this simulation, the 290
Simulink Look-up table skill is employed to simulate the reactivity change using a time 291
series reactivity table. Two tables representing the internal and external reactivity change 292
are represented in Table 2 and Table 3. 293
10
294
Table 2 Reactivity change due to insertion of shutdown rods [19] 295
296
Time Reactivity (mk)
0.0 0.0
0.4 -2.0
0.6 -5.46
0.8 -9.10
1.0 -13.64
1.2 -30.28
1.4 -47.36
1.6 -64.24
1.8 -80
297
Table 3 Internal reactivity change with time [1] 298
Time Reactivity (mk)
0.0 0.0
0.9 4.3
3.16 11
10 15
299
Then, the reactor dynamic model is simulated by MATLAB/Simulink. Fig. 2 represents the 300
simulation model for CANDU reactor LLOCA accident. Powers transient simulation 301
results with and without the reactor trip are shown in Fig. 3. It is basically consistent with 302
the power transient curve revealed by the COG report [1], except that the peak power is a 303
bit lower and the downstream inflection point shows a bit later. This explains that the 304
approximated internal reactivity table still cannot accurately reflect the real situation in the 305
reactor core when large LOCA happens. Another potential factor is that, the COG report 306
shows the peak bundle power transient, which is essentially different from the reactor bulk 307
power. 308
309
11
s
1
[ ]'i
( )jdiag
n
][ iC
3
1
1 10
s
1
Reactivity Table 2.2
Reactivity Table 2.3
3
1( )(1 ) 11 10
i
internal
external
[ ]i
1
310
311
Fig. 2 Block diagram of the reactor model using a point kinetic method 312
313
Fig. 3 Power transient of CANDU reactor LLOCA event (with and without reactor 314
trip) 315
316
3. SDS1 trip logic simulation for the LLOCA event 317
With manipulation of 28 mechanical shutoff rods, SDS1 is the preferred shutdown system 318
to quickly terminate reactor operation when certain parameters exceed specified limits. 319
This preference is with the economic consideration resulting from plant unavailability 320
following the use of SDS2. SDS1 employs an independent triplicate logic system, which 321
senses the requirement for reactor trip and de-energizes the direct current clutches to release 322
0 1 2 3 4 5 60
0.5
1
1.5
2
2.5
3
3.5
4
Time (Sec)
Norm
aliz
ed r
eacto
r p
ow
er
(FP
U)
without trip
with trip
12
the spring assisted gravity drop shutoff rods. 323
Fig. 4 represents the brief structure of one of the three CANDU SDS1 channels, which 324
consists of sensors for system variable measurement, trip computer for trip logic processing, 325
relay logic for 2oo3 voting, and the shutoff rods for reactor trip [20]. 326
327
Sensors Amplifier PDC(Trip Computer)
Conditioning
Signals
Relay
Shutoff Rods
Monitoring
Computer
Sampling Delay Decision Making Relay delay Rods Dropping
Display/Test
ComputerTest Signals
Gain
328
329
Fig. 4 Signal architecture of SDS1 330
331
From the above SDS1 signal architecture figure it can be seen that, the time spent by a 332
shutdown process is the summation of the time consumed in each of the four sections. The 333
shutdown process can be speeded up by reducing the consumed time of any of these parts. 334
In this research, the attempt is focused on the first two parts, i.e. reducing the consumed 335
time on sensor sampling delay and the trip decision making during the LLOCA accident. 336
In order to achieve this objective, decomposition of these two blocks is performed. 337
Design objectives of the shutdown systems are obtained by considering which mechanisms 338
may lead to violation of the derived criteria and by setting conservative objectives to meet 339
the criteria. The considerations are expressed in terms of design basis initiating events. 340
LOCA is one of the main designed basis events. The design objective for LLOCA is to 341
maintain the integrity of the primary heat transport system. This requires that the integrity 342
of fuel channels be maintained. Channel integrity at high pressure is maintained providing 343
that fuel elements do not contact a pressure tube. For LLOCAanalysis the prevention of 344
fuel breakup is chosen as the trip effectiveness criterion for SDS1. 345
The selection of trip parameters is such that there are adequate measurements for all process 346
failures identified. Trip parameters, setpoints, and protective coverage of LLOCA are 347
shown in Table 4. 348
349
13
Table 4 SDS1 Trip Parameters and Setpoints for LLOCA 350
351
No. Trip Parameters Detector type Setpoint
1 High Neutron Power Vertical in-core detectors
ROPT-HSP-1 122% FP
ROPT-HSP-2 109.3% FP
ROPT-HSP-3 83.3% FP
2 High Log Rate
Neutron Power Ion chambers 10% pp/sec
352
The overpower trip provides a quickly responding trip signal for LLOCA accidents where 353
the induced void reactivity rate or depth exceeds the capability of the reactor regulating 354
system to maintain power constant. The high log rate power trip is also designed to give 355
protection against the LLOCA event. A trip setpoint of 10% present power per second (pp/s) 356
with a system response time of less than 1 second is required. This can be achieved with a 357
second-order rate filter with time constants of approximately 0.16 seconds. The high log 358
rate power trip dominates the LLOCA event, which means its trip time is faster than the 359
overpower trip. 360
Reactor flux power measurements for the regional overpower trip are provided by Pt-clad 361
in-core flux detectors, located in vertical in-core flux monitoring assemblies. Reactor 362
power measurements for the log rate neutron power trips are provided by ion chambers, 363
located on the inaccessible side of the calandria. The Pt-clad in-core detectors are 364
characterized with a response which is approximately 90% prompt to neutron flux. The 365
overall response is a good representation of the power-to-fuel dynamic characteristics of a 366
CANDU reactor. An amplifier converts the detector current to a suitable voltage range. 367
This voltage signal is compensated to account for the delayed component of the detector 368
signal. The time constants of this compensation are adjusted so that the output signal 369
closely matches the fuel power. Three uncompensated ion chambers are provided to 370
measure the log rate reactor power for SDS1. The output current from each ion chamber 371
goes to an amplifier, which produces log neutron power, linear neutron power, and log rate 372
signals. The log rate signal is a direct trip parameter. The log and linear power signals are 373
used for conditioning and trip setpoint selection for other trip parameters. 374
Since the high log rate trip dominates the LLOCA event, it is focused on analysis in this 375
research. The decomposition of the trip initiation delays in instrumentation and estimated 376
measurement is performed. For this trip signal, the log power signal is first differentiated 377
and then smoothed by two first order filters with a delayed time constant. The filtering is 378
used because the result of differentiating a noisy signal is a very noisy signal. The pure 379
delay resulting from the second order filter depends on the characteristics of the input. The 380
relay delay, clutch delay, and so on, will be the same as for the high power trip, which is 381
not discussed in this research. The time constant of the log amplifier, which is negligible at 382
14
high powers, is 40 milliseconds. This time constant is usually modeled as a function of 383
power, to account for its increase in magnitude at very low powers. 384
A transfer function flow chart for high log rate trip logic of LLOCA event is shown in Fig. 385
5, where T1 is the delay time constant of the amplifier, and T2 is the delay time constant of 386
the filter. 387
Trip
setpoint
Neutron
fluxln Comparator
Trip
signal1
1
1 T s
du
dt2
2
1
(1 )T s 388
389
Fig. 5 Transfer function chart of high log rate trip logic for LLOCA event 390
391
Reactor power transient includes three types of noises: sine wave of 60 Hz, 1000 Hz, and 392
random noise of 200 Hz [21]. The amplitude of the noises is 0.5%FPU. Trip setpoint is 393
designated by 10% pp/s. If the induced signal is over 10%pp/s, the output signal will be on 394
1; otherwise, the output will be still on 0. The simulation result of the trip signal is provided 395
in Fig. 6. From the picture it can be shown that the reactor high log rate trip happens at 396
222.3 milliseconds. 397
398
Fig. 6 Simulation result of trip signal for existing trip system 399
400
0 0.2 0.4 0.6 0.8 10
1
Time (Sec)
Tri
p s
ign
al
15
4. The proposed new signal processing system 401
4.1 Schematic presentation of the new signal processing system 402
To address the poor signal-to-noise ratio of neutronic power measurement signals, 403
especially at low reactor power, and insufficient speed of trip signal generation in case of 404
LLOCA, a new neutronic signal processing system is developed. An improved relative rate 405
trip concept is provided that it neither causes significant delays nor has substantial signal-406
to-noise ratio restrictions. Schematic representation of the system and trip logic is shown 407
below in Fig. 7. 408
Neutron
Flux
Detector(s)
Amplifiers
Variable Set
Point
FRate
ModuleF
Signal
Processing
Unit
F
Tri
p L
og
ic
Legend: F - filter 409
Fig. 7 Simplified schematic representation of the new neutronic signal processing 410
system for LLOCA 411
412
A test to low pass filters following a power pulse transient is performed by Simulink. Three 413
first-order low pass filters are provided with delay time constants of 5 second, 50 414
milliseconds, and 20 milliseconds, respectively. Power transient and three filters’ transient 415
characteristics are represented by Fig. 8. It can be seen from the figures that, during the 416
power transient process, the filter with a longer delay time constant represents a relatively 417
prolonged time characteristics, although all of the filters’ responses will arrive at the 418
designated power level eventually. This is coincident with the anticipation. 419
16
420
Fig. 8 Power step response for first-order low pass filters with different time 421
constants ((a) –power step trajectory; (b) – 5 seconds; (c) – 50 milliseconds; (d) – 20 422
milliseconds) 423
424
4.2 Software design and implementation of the new signal processing 425
system 426
Inspired by the above power step response characteristics of different first-order low pass 427
filters, a specified design of the new signal processing system corresponding to Fig. 7 is 428
provided. Fig. 9 represents the transfer function blocks of the new designed system’s 429
principles. Again, three first-order low pass filters are used. Their time constants are 430
respectively 5 seconds, 50 milliseconds, and 20 milliseconds. 431
432
0 5 10 15 200
0.2
0.4
0.6
0.8
1
Time (Sec)
Pow
er
(FP
U)
(a)
0 5 10 15 200
0.2
0.4
0.6
0.8
1
Time (Sec)
Pow
er
(FP
U)
(b)
0 5 10 15 200
0.2
0.4
0.6
0.8
1
Time (Sec)
Pow
er
(FP
U)
(c)
0 5 10 15 200
0.2
0.4
0.6
0.8
1
Time (Sec)
Pow
er
(FP
U)
(d)
17
Trip
setpoint
Neutron
fluxln
Comparator
Trip
signal
1
1 Ts
2
1
1 T s
1
1
1 T s
3
1
1 T s
Proportioner
Comparator
Comparator
Proportioner Comparator
Trip
setpoint
.AND.
433
434
Fig. 9 Transfer function chart of new designed signal processing system for LLOCA 435
event (T1, T2, and T3 are respectively 5 seconds, 50 milliseconds, and 20 436
milliseconds) 437
438
A MATLAB/Simulink simulation platform is established to model the new designed signal 439
processing system. Power transient is induced from the theoretical analysis which is 440
illustrated in Section 2.2. Considering the delay effect of filters particularly as the proposed 441
low pass filter with 5 second time delay, the response of the filters is observed. It takes 442
more than 50 seconds for the 5 second low pass filter to reach a new steady state level. 443
Therefore, in order to maintain the initial condition of full power steady state operation 444
before the LLOCA event is issued, full power operation steady states of 60 seconds are pre-445
arranged. 446
Reactor power transient includes three types of noises: sine wave of 60 Hz, 1000 Hz, and 447
random noise of 200 Hz. The amplitude of the noises is 0.5%FPU. Three low pass filters 448
with time constant of 5 s, 50 ms, and 20 ms are used to process the power transient signal. 449
Since the response of 5 s filter is much lower than other two filters, it can be taken into 450
account as a constant value in a very short time. Therefore, in the same time period, 451
equations of (F2-F1)/F1 and (F3-F1)/F1, where F1, F2, and F3 represent the response 452
values of 5 s, 50 ms, and 20 ms filters respectively, can be used to measure the relative 453
deviations between fast filters and the 5s filter. Based on these measurements, an internal 454
setpoint corresponding to the original setpint, 10 pp/s can be designed, as far as the 455
proposed trip time reducing objective is concerned. 456
Simulation results about the reactor trip signals respect to both the existing and new 457
designed systems are represented in Fig. 10. The reactor trip time under the action of the 458
new designed signal processing system is 138 ms, which is significantly shorter than the 459
old one, 222 ms. Furthermore, no spurious trips happen to both trip systems. Therefore, the 460
conclusion is that, the new designed signal processing system achieve the faster trip logic 461
in the same trip setpoint condition as the existing design, and meanwhile suppress the 462
18
spurious trips happening. Subsequently, the new designed systems will significantly 463
enhance the power plant operation safety margin. 464
465
Fig. 10 Simulation results of reactor trip signals for existing and new trip systems 466
467
4.3 Discussions and analysis 468
The new neutronic signal processing system will bring the economic benefit to the 469
industrial utilizations by reducing the reactor trip time during the LLOCA event. 470
Conservatively assuming the reactor trip time is decreased by 80 ms which is caused by 471
implementation of the new signal detection system, the reactor peak power is reduced from 472
3.06 FPU to 2.695 FPU during the LLOCA event. Then the reactor operation safety margin 473
is significantly increased. Fig. 11 represents the power transient curves corresponding to 474
the existing and new systems during the LLOCA event. If the reactor operation safety 475
margin is maintained same as before, the reactor operation power level could be 476
considerably increased. For example, through a “core conversion” procedure, Bruce Power 477
B units have been increased from 0.90 FPU to 0.93 FPU. Combining all four Bruce B units, 478
this is a 100 megawatt increase in generation, enough electricity to power 100,000 Ontario 479
homes [22]. Furthermore, with the application of the new designed system, the potential 480
power operation level of Bruce B units could be very close to 1.0 FPU. This is an extra 200 481
megawatt increase in generation, which is equal to an economic benefit of about 0.2 million 482
dollars per hour. 483
484
0 0.2 0.4 0.6 0.8 10
1
Time (Sec)
Tri
p s
ign
al
existing system
new system
19
485
Fig. 11 Estimation of the reactor peak power decrease during the LLOCA event 486
caused by the new signal processing system 487
488
5. Conclusion 489
An innovative neutronic signal processing system for CANDU reactor SDS1 manipulation 490
is proposed, and its detailed design based on software simulations is performed. 491
Implementation of the new signal processing system represents a shorter trip time during 492
the LLOCA event than the existing power plant system. Furthermore, the proposed 493
algorithms and their implementation are proved to be able to accommodate both trip 494
effectiveness and spurious trip immunity. Implementation of the new designed signal 495
processing system will provide enhanced safety margin for the existing power plant 496
operation; or bring extra economical benefits for the units by increasing the current 497
operation power level when the original safety margin is maintained. 498
Acknowledgments 499
The authors would like to acknowledge the financial support from the NSERC for the work 500
reported in this paper. 501
502
503
0 1 2 3 4 50
0.5
1
1.5
2
2.5
2.6950
33.0649
3.5
Time (Sec)
Rea
ctor
Pow
er
(FP
U)
Existing system
New system
20
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