Nuclear Instrumentation

Abstract

Instrumentation for reactor control and protection needs to cover a wide range of measurement targets. In addition to neutron flux measurements for monitoring reactor power, a variety of measurements is required, such as coolant temperature, pressure and flow rate measurements, depending on reactor power and type. These measurements can be roughly divided into two categories. One is nuclear instrumentation, which measures neutrons and $\gamma$-rays, and the other is process instrumentation, which is used for general industrial measurements such as temperature and pressure measurements. In a zero-power reactor such as UTR-KINKI, nuclear instrumentation plays a crucial role. This chapter describes characteristics and compositions of nuclear instrumentation, including radiation detectors used for UTR-KINKI as an example.

5.1 Introduction

A variety of information is necessary for the safe and stable operation of a nuclear reactor. In addition to measuring neutrons in the reactor to monitor the reactor power, more information is necessary on the temperature, pressure, flow rate and water level of the coolant in a high-power reactor. Other measurements include those for detecting impurities in the coolant and fuel failure, and there are measurement targets and techniques. Although the classification of these measurements is not strictly defined, the instrumentation used in general industrial measurements that does not involve measurements of neutrons and $\gamma$-rays is called the process instrumentation (non-nuclear instrumentation), and the instrumentation specific to nuclear technology, which measures neutrons and $\gamma$-rays, is called the nuclear instrumentation.

For a zero-power reactor such as UTR-KINKI, nuclear instrumentation mostly satisfies requirements for the control of the reactor. In UTR-KINKI, information on reactor power is obtained by five neutron detectors installed on the reflector of the reactor. The process instrumentation installed in UTR-KINKI consists of a seismic detector, a moderator thermometer and the water level gauge of the biological shield tank. In this chapter, we focus on nuclear instrumentation and introduce neutron detectors used for nuclear instrumentation, and then explain the nuclear instrumentation of UTR-KINKI.

Power reactors have special neutron detectors installed in the reactor core (in-core detectors) in addition to out-core detectors to obtain information such as spatial distribution of neutron flux. Process instrumentation also plays an important role in power reactors, and readers who want to know about these measurements are advised to see the references at the end of the chapter.

5.2 Nuclear Instrumentation

In the operation of nuclear reactors, the highest priority is to ensure safety in all processes from start-up to shutdown of the reactor. Moreover, the reliability of the measurement system is most required for nuclear instrumentation, unlike that for general industrial measurement or experimental research. Neutron detectors for nuclear instrumentation are chosen to be simple in structure and detection principle and to have few failures, rather than highly accurate and complicated detectors. In addition, redundancy should be provided especially for detector systems that are directly related to safety, and detectors with the same function should be used in several independent channels simultaneously to prevent complete loss of function.

Another characteristic of nuclear instrumentation is that the measurement range is extremely wide compared with general radiation measurements. In general, the neutron flux in a nuclear reactor varies with about eight to ten orders of magnitude during the period from start-up to operation at a full power, and one type of neutron detector cannot cover the entire measurement range. Then, the power range of the reactor is usually divided into three ranges and the information from the detectors in charge of each range is combined to obtain continuous information over the whole power range. In addition, the measurement of each range is made to overlap each other so that there is no discontinuity in the measurement when moving from one range to the other. The power range of a reactor is generally divided as follows:

• (1) Startup range

The range in which a neutron source is inserted into a reactor to start up the operation. It is also called the source range.

• (2) Intermediate range

The range in which the control rods are withdrawn and the power of the reactor is gradually increased. The exponentially increasing power is monitored.

• (3) Power range

The range in which the reactor is operated at a full power ranging between 1 and 100%. Automatic operation is performed in this range.

The corresponding neutron flux to these power ranges depends on the full power of the reactor. Figure 5.1 shows a conceptual diagram of the power ranges. The source range and the intermediate range are sometimes referred to together as the startup range.

Fig. 5.1

General power range of a research reactor

5.3 Neutron Detectors for Nuclear Instrumentation

The thermal power (W) of a nuclear reactor is the sum of the energy generated per unit time (J s⁻¹) from fission reactions in the reactor. Since fission reactions are triggered by neutrons in the reactor, the thermal power of the reactor is proportional to the number of neutrons in the reactor, which is the neutron flux as a specific physical quantity. The thermal power of the reactor can be then monitored by installing a neutron detector at a position where a sufficient number of neutrons proportional to the average neutron flux in the reactor are incident and detecting them. The obtained information on the neutron flux is used not only for reactor operation control but also for safety by producing scram signals. Since fission reactions are triggered by thermal neutrons in thermal neutron reactors, thermal neutron detectors are used for the nuclear instrumentation.

When various types of thermal neutron detectors have been developed and used for general radiation measurement applications, gaseous detectors are mainly used for nuclear instrumentation. This is because existing gaseous detectors for thermal neutron measurement are suitable for nuclear instrumentation for their simple structure, high reliability, stable operation for a long period of time, wide operating range, excellent n-$\upgamma$ discrimination performance, resistance to radiation damage, and low susceptibility to activation by neutrons. Research and development of scintillation and semiconductor detectors for neutron measurement continue, but they are not used for nuclear instrumentation because many scintillation detectors are sensitive to $\upgamma$-rays and false signals from photomultiplier tubes cannot be ignored under a high radiation environment, and semiconductor detectors are vulnerable to radiation damage.

Since the object of measurement is a quantity proportional to the neutron flux, information on the energy of neutrons is not necessary, and information on the number of neutrons incident on the detector is sufficient.

5.3.1 BF-3 Proportional Counters

The detector that is filled with boron trifluoride (BF-3) gas as a detector gas is called the BF-3 proportional counter. Figure 5.2 shows the structure of a typical BF-3 proportional counter. Since neutrons do not have an electric charge, they cannot ionize the detector gas directly. An isotope of boron, B-10 (natural isotope abundance: 19.9%), has a large (n, $\mathrm{\alpha }$) reaction cross section for thermal neutrons (3837 barns for 0.0253 eV neutron), and the following charged-particle production reaction occurs:

$${}^{10}{\text{B}} + {\text{n}} \to {}^{{7}}{\text{Li}} + {}^{4}{\text{He}}{.}$$

Fig. 5.2

Structure of BF₃ proportional counter

In this reaction, 94% of the produced lithium-7 (Li-7) nuclei are left in the excited state of 0.482 MeV, and 6% are in the ground state. By using this reaction, thermal neutrons incident on the BF-3 proportional counter react with B-10 in the gas and are converted into two charged particles, Li-7 and helium-4 (He-4) nuclei, which ionize the gas. It is the same principle as that of the ordinary proportional counter that a signal is generated in the process of collecting electrons and ions generated in the gas at the electrodes. In other words, the BF-3 gas plays two roles of converting thermal neutrons into charged particles and of acting as a detector gas of the proportional counter.

The Q-value of the $^{{{1}0}} {\text{B}}(n,\alpha )^{{7}} {\text{Li}}$ reaction is 2.310 MeV when Li-7 is left in the excited state and 2.792 MeV when it is left in the ground state. Since the energy of thermal neutrons is almost negligible: much smaller than 2.310 and 2.792 MeV, the sum of kinetic energies of Li-7 and He-4 is 2.310 MeV or 2.792 MeV, which is given to the BF-3 gas. This means that the output signal of the BF-3 proportional counter does not reflect the incident neutron energy, but only the Q-value of the reaction. Since the signal originated from the Q-value is larger than that from $\upgamma$-rays or electronic noise, it is easy to discriminate only the signals caused by the neutrons.

The BF-3 proportional counter is usually a tube made of a metal cylinder with a thin wire stretched in the center. A high voltage is applied between the wire (anode) and the tube wall (cathode), and pulse signals are generated by incident thermal neutrons. The sensitivity depends on the volume of the tube and the gas pressure. Since the BF-3 proportional counter is generally highly sensitive, it is mainly used in a pulse mode in a startup range.

[Example] Calculate the kinetic energies of the Li-7 and He-4 nuclei produced by the $^{{{1}0}} {\text{B}}(n,\alpha )^{{7}} {\text{Li}}$ reaction, assuming that the Li-7 nucleus produced by the reaction is left in the excited state, and the Q-value of the reaction is 2.310 MeV.

Since the energy of thermal neutrons ($\sim$ 0.0253 eV) is almost negligible compared to the Q-value of the reaction, the B-10 nucleus and the thermal neutron can be considered to have been at rest in the initial state, and at some point they reacted, emitting Li and He nuclei with large kinetic energy in opposite directions. The reason why the two nuclei are emitted in opposite directions is that the sum of the momentum of thermal neutrons and the B-10 nucleus is zero due to the conservation of momentum, since they are originally at rest.

If the masses of the Li-7 and He-4 nuclei are ${\mathrm{m}}_{\mathrm{Li}}$ and ${\mathrm{m}}_{\mathrm{\alpha }}$, and the kinetic energies are ${\text{E}}_{{{\text{Li}}}}$ and ${\text{E}}_{\alpha }$, respectively, the following relations can be obtained from the conservations of energy and momentum:

$${\text{E}}_{{{\text{Li}}}} + {\text{E}}_{\alpha } = {\text{Q}},$$

(5.1)

$$\sqrt {2m_{{{\text{Li}}}} {\text{E}}_{{{\text{Li}}}} } = \sqrt {2m_{\alpha } {\text{E}}_{\alpha } } .$$

(5.2)

From Eq. (5.2), we get

$$\frac{{{\text{E}}_{{{\text{Li}}}} }}{{{\text{E}}_{\alpha } }} = \frac{{m_{\alpha } }}{{m_{{{\text{Li}}}} }}.$$

(5.3)

From Eq. (5.3), the ratio of ${\text{E}}_{{{\text{Li}}}}$ to ${\text{E}}_{\alpha }$ is the inverse ratio of the masses of Li-7 and He-4 nuclei. As the ratio of ${\text{m}}_{{{\text{Li}}}}$ to $m_{\alpha }$ can be regarded as 7/4 from their mass numbers, the ratio of ${\text{E}}_{{{\text{Li}}}}$ to ${\text{E}}_{\alpha }$ is 4/7.

Therefore, we obtain the energies of ⁷Li and ⁴He nuclei as follows:

$${\text{E}}_{{{\text{Li}}}} = \frac{{m_{\alpha } }}{{m_{{{\text{Li}}}} + m_{\alpha } }}Q \approx \frac{4}{7 + 4} \times 2.310 = 0.840\;{\text{MeV}},$$

(5.4)

$${\text{E}}_{\alpha } = \frac{{m_{{{\text{Li}}}} }}{{m_{{{\text{Li}}}} + m_{\alpha } }}Q \approx \frac{7}{7 + 4} \times 2.310 = 1.470\;{\text{MeV}}.$$

(5.5)

[Column] Pulse height spectrum of BF-3 proportional counter

When the output from a BF-3 proportional counter is analyzed with a multichannel analyzer (MCA), a pulse height spectrum will have a complicated structure as shown in Fig. 5.3. This characteristic structure of the pulse height spectrum can be explained by a phenomenon called the wall effect of the BF-3 proportional counter.

Fig. 5.3

Pulse height spectrum of BF-3 proportional counter

What would the pulse height spectrum look like if we had a very large BF-3 proportional counter? In this case, the total energy of He-4 and Li-7 produced in $^{{{1}0}} {\text{B}}(n,\alpha )^{{7}} {\text{Li}}$ reactions would be fully absorbed in the gas, and the pulse height spectrum should show two peaks at the channels corresponding to 2.79 MeV when Li-7 is in the ground state (6%) and 2.31 MeV when it is in the excited state (94%), as shown in Fig. 5.4. The difference in the height of the peaks reflects the probability that Li-7 populates in the ground state or excited state.

Fig. 5.4

Ideal pulse height spectrum from a large BF-3 proportional counter with 2.31 MeV (He-4) and 2.79 MeV (Li-7) nuclei fully absorbed in the gas

The diameter of the BF-3 proportional counter is typically on the order of centimeters, and the volume of the counter is not large compared to the range of He-4 and Li-7 nuclei produced in the gas. Then, if the reaction occurs near the inner wall of the counter, either He-4 or Li-7 will collide with the wall and be absorbed (remember that He-4 and Li-7 are always emitted in opposite directions from the conservation law of momentum), and a part of the energy may not be deposited in the gas. Let us assume that Li-7 is left in the excited state, and focus on the He-4 produced in the reaction occurred near the inner wall. As shown in Fig. 5.5a, if the reaction occurs on the inner surface of the tube, Li-7 deposits all its energy in the gas while He-4 is absorbed by the wall without depositing its energy. In this case, the only energy deposited in the gas is the energy of Li-7, which means that the energy 0.84 MeV is deposited in the gas. Next, as shown in Fig. 5.5b, if the reaction occurs at a distance anywhere within the range of He-4 from the inner wall surface, all of the energy of Li-7 and part of the energy of He-4 will be absorbed in the gas. As shown in Fig. 5.5c, when the reaction occurs at a distance larger than the range of He-4 from the inner wall surface, the total energy of Li-7 and He-4, 2.31 MeV, is absorbed in the gas. Therefore, the pulse height spectrum produced by the reactions that occur at a distance within the range of He-4 from the inner wall will be a uniformly distributed rectangle spectrum from 0.84 MeV to 2.31 MeV, as shown in Fig. 5.6. As shown in Fig. 5.7, the pulse height spectrum from the reactions that occur at a distance within the range of Li-7 will also be a uniformly distributed rectangle spectrum from 1.47 MeV to 2.31 MeV.

Fig. 5.5

He-4 generated from the reaction near the inner wall of the BF-3 proportional counter absorbed by the tube wall

Fig. 5.6

Pulse height spectrum produced from the reactions where He-4 is absorbed by the wall

Fig. 5.7

Pulse height spectrum produced from the reactions where Li-7 is absorbed by the wall

As a result, a pulse height spectrum from a BF-3 proportional counter is a superposition of the following three spectra:

• (1) The peak that is produced when all the energies of both Li-7 and He-4 are absorbed in the gas

• (2) The rectangular distribution that is produced when the part of the energy of He-4 is absorbed in the inner wall

• (3) The rectangular distribution that is produced when the part of the energy of Li-7 is absorbed in the inner wall.

The three spectra are produced for the ground state and the excited state of Li-7, and the resulting spectrum will have the structure shown in Fig. 5.3.

5.3.2 Boron-Lined Proportional Counters

The boron-lined proportional counter is a proportional counter whose inner wall is coated with B-10-enriched boron. In the same manner as BF-3 proportional counters, the $^{{{1}0}} {\text{B}}(n,\alpha )^{{7}} {\text{Li}}$ reaction by incident thermal neutrons is used to produce charged particles, ⁷Li and ⁴He, to ionize the gas. Unlike the BF-3 proportional counter, only one of the two charged particles can deposit its energy in the gas because the products are emitted oppositely. When one product is emitted into the gas, the other product is always absorbed in the coating layer and the wall. The efficiency of boron-lined counters can be improved as the thickness of the boron-coated layer is increased. The thickness, however, cannot be thicker than the range of the product because the products produced in the layer farther than the range from the inner surface cannot reach the gas and are never detected. Therefore, the maximum thickness of the coating will be determined by the range of ⁴He, which is on the order of 1 mg ⋅ cm⁻². The advantage of boron-lined proportional counters over BF-3 proportional counters is that it can use more suitable proportional gas than BF-3 gas. Boron-lined proportional counters are also used in pulse mode in a startup range. Figure 5.8 shows the structure of a typical boron-lined proportional counter.

Fig. 5.8

Structure of boron-lined proportional counter

5.3.3 Fission Counters

The fission counter (FC) is a gaseous detector that converts thermal neutrons into charged particles using fission reactions. In other words, thermal neutrons are detected when fission fragments produced in fission reactions ionize the detector gas.

The surface of the electrodes of FC is coated with U₃O₈ enriched with U-235, and argon (Ar) is used as the detector gas. The energy generated by fission reactions of U-235 is quite large, about 200 MeV, of which about 170 MeV is the kinetic energy of fission fragments. Since the energy of fission fragments is large and the specific ionization (the number of ion pairs produced per unit path when a charged particle travels through a material) is also large, a sufficiently large charge can be obtained without gas amplification, and the detector can be operated as an ionization chamber.

As in the case of boron-lined proportional counters described in Sect. 5.3.2, since two fission fragments are also directed oppositely, when one fission fragment ionizes the gas, the other is absorbed in the coating layer. Then, only one of the two fission fragments will always impart energy to the gas. If the thickness of the coating layer is increased to gain a high efficiency, the fraction of the energy absorbed in the coating layer increases, and the average energy imparted to the gas decreases.

Since U-235 and U-238, which are isotopes of uranium, are alpha emitting radioisotopes, alpha particles are always emitted from the coating layer and causes background events. Meanwhile, the energy of fission fragments is more than ten times larger than that of alpha particles, while the energy of alpha particles is about 5 MeV. Therefore, it is easy to discriminate the pulse height based on the difference in the magnitude of the signals, and, as a result, only the signals from neutrons can be counted.

Although the efficiency of FC is generally inferior to that of BF-3 proportional counters, FC is suitable for measurements at high counting rates because the rise time of the signal is short, ranging between 0.1 and 0.3 $\upmu$ s.

5.3.4 Ionization Chambers

The ionization chamber is the simplest detector in terms of detection principle and structure among various radiation detectors, and is suitable for nuclear instrumentation from the viewpoint of reliability. In ranges such as the intermediate range and the power range, where neutron flux is larger and counting rate is too high to measure in the pulse mode, the ionization chamber is used in the current mode. In the ionization chamber for neutron measurement, the B-10-enriched boron is deposited on the electrodes, and the He-4 and Li-7 nuclei emitted from the $^{{{1}0}} {\text{B}}(n,\alpha )^{{7}} {\text{Li}}$ reactions ionize the gas.

Since the ionization chamber for nuclear instrumentation is used in an environment where neutrons and $\upgamma$-rays are mixed, both neutrons and $\upgamma$-rays produce electric charges in the chamber. When used in the current mode, neutrons and $\upgamma$-rays contribute to the output current because neutrons and $\upgamma$-rays cannot be discriminated by pulse height unlike in the case of pulse mode detectors.

The compensated ionization chamber (CIC) is the ionization chamber that has a mechanism to remove the contribution of $\upgamma$-rays from the output current. The principle of the operation is shown in Fig. 5.9. The CIC consists of two ionization chambers sharing one electrode, and only the electrode of one of two chambers is coated with the enriched B-10. The two chambers are sensitive to $\upgamma$-rays, whereas one of them is also sensitive to neutrons. By adjusting the sensitivity of two chambers to $\upgamma$-rays to be the same, only the current from the neutron contribution can be extracted.

Fig. 5.9

Operation principle of compensated ionization chamber (CIC)

In an environment where the contribution of $\upgamma$-rays is small and negligible, or where reliability is more important than accuracy, such as detectors used for a safety channel, the uncompensated ionization chamber (UIC) is then used. UIC is an ionization chamber with a simple structure whose electrodes are coated with boron to make it sensitive to neutrons.

5.4 Compositions of Nuclear Instrumentation

Neutron detectors used in nuclear instrumentation are classified into four channels according to the power range to be monitored and roles: (1) startup channel, (2) intermediate channel, (3) power channel and (4) safety channel.

In this section, the power range to be monitored by the nuclear instrumentation of UTR-KINKI is introduced, as shown in Fig. 5.10, and compositions of the nuclear instrumentation are presented, as shown in Fig. 5.11. Moreover, in the following, the nuclear instrumentation is explained by using several channels used in UTR-KINKI as an example.

Fig. 5.10

Measurement power range of nuclear instrumentation of UTR-KINKI

Fig. 5.11

Compositions of nuclear instrumentation of UTR-KINKI

5.4.1 Startup Channel

Startup channel is used in the startup range where a neutron source is inserted into the core to start up the reactor. The BF-3 proportional counter, the boron-lined proportional counter or FC may be used for the startup channel. Since the neutron flux is sufficiently small in the startup range, any of these detectors can be used in the pulse mode. The signals from the detectors are amplified and shaped by a preamplifier and a linear amplifier, and the signals caused by $\upgamma$-rays and electronic noise are then removed by a pulse height discriminator, and, as a result, neutron signals are counted and displayed.

The FC is used as the startup channel of UTR-KINKI, and the neutron counting rate (cps) is displayed on the digital rate meter in the control console. The signals from the startup channel is also used as one of the conditions for the startup sequence of the reactor. When the counting rate is lower than 10 cps, the withdrawal of control rods is prohibited as “Low Count Rate”.

5.4.2 Intermediate Channel

Intermediate channel is used to measure the power and reactor period in the process of increasing the reactor power by withdrawing the control rods, and the CIC is used as neutron detectors in the current mode. In the intermediate channel, the power, which varies exponentially over a wide range from the startup range to the power range, needs to be monitored on a single scale. For this purpose, the current signal from the CIC is converted into a voltage signal proportional to the logarithm of the input current by a logarithmic amplifier and displayed on a logarithmic power meter (Log-N meter).

The output from the logarithmic amplifier is also used for a period meter. When the power of the reactor increases exponentially, neutron flux n increases as

$$n = n_{0} e^{\frac{t}{T}} ,$$

(5.6)

where T is the reactor period and is defined as the time required to change the power by a factor of e. When the signal proportional to n is input to a logarithmic amplifier, the output is obtained by taking the logarithm of both sides in Eq. (5.6),

$$\ln \;n = \ln \;n_{0} + \frac{t}{T}.$$

(5.7)

Differentiating Eq. (5.7) gives

$$\frac{d}{dt}\left( {\ln \;n} \right) = \frac{1}{T},$$

(5.8)

and the period T is obtained from this relationship. If the output signal from the logarithmic amplifier is input to a differentiation circuit, and the output signal from the differentiation circuit is displayed on a reciprocal scale, the display indicates then a period meter.

The CIC is used for the intermediate channel of UTR-KINKI, and the output current (A) and reactor period (s) are displayed on the logarithmic power meter (Log-N meter) and period meter of the control console, respectively. The output from the period meter is also used to trigger the signals of alarm and scram. When the period is shorter than 10 s, alarm is triggered and the buzzer sounds, and when the period is shorter than 5 s, a scram signal is generated and the reactor is shut down immediately.

5.4.3 Power Channel

Power channel is used to monitor the reactor power ranging between 1 and 100%, and is also used for the automatic control of operation to maintain the power at the demand level. The CIC or UIC is used as a neutron detector in the current mode. The output current from the chamber is amplified by a linear amplifier and converted to a current signal or a voltage signal proportional to the input current, which is generally displayed on a linear power meter (Lin-N meter).

UTR-KINKI uses the CIC as a neutron detector for the power channel, and, unlike the general configuration described above, the output current from the CIC is directly measured by a picoammeter (an ampere meter for small current measurement) and is displayed on the linear power meter. The output current is continuously measured in the entire range from the startup range to the power range to monitor the power. Since the output current varies greatly over five decades ranging between 10^–11 and 10⁻⁷A, the measurement is performed by switching the measurement range as the power increases. The relationship between the readings of the linear power meter and the power of the reactor is calibrated in advance, and the power of the reactor is 1 W when the output current is 5.72 × 10^–8 A.

The signal from the linear power meter is input to the power recorder (chart recorder), which constantly records the change in the reactor power on a chart paper during the reactor operation. When the reading of the linear power meter exceeds the measurement range, and the indication of the recorder is out of the display range (over-range), an alarm is triggered as “power recorder off scale.” Furthermore, the signal from the power recorder is used for automatic control of reactor operation, and is input to the servo controller for automatic operation. The servo controller is equipped with a switch to select automatic or manual operation. When the switch is selected to the automatic operation, the regulating rod (RR) is positioned by the proportional–integral–deviate (PID) control, and automatic operation is performed to maintain the reactor power at the demand level. The deviation (%) between the demand levels set by the servo controller and the actual reactor power is displayed on the servo deviation meter. Here, if the deviation exceeds 10% during the automatic operation, an alarm is triggered as “servo deviation over 10%”.

5.4.4 Safety Channel

Safety channel monitors the abnormal increase of neutron flux for reactor protection. Neutron detectors for safety channel are required to have high reliability and quick response rather than accuracy. To improve the reliability, several independent detectors are used simultaneously, and the system is multiplexed so that the function is not completely lost even if a failure is happened in one of the detectors.

The safest approach for safety channel is to scram a reactor whenever an anomaly is detected, even if it is attributable to a false detection. This kind of approach is called “fail safe,” which is adopted for low-power reactors that are easy to start up and shut down. On the other hand, in a nuclear power plant or a high-power research reactor, if the reactor is shut down due to a false detection, it takes a long time to restart, resulting in a significant loss of availability of the reactor, significant economic and social impact. Therefore, it is important to keep the reliability of the safety channel high. A typical example of such a system is the “2 out of 3” system that three detector systems are prepared for the same measurement target. Of three systems, when only one system detects an anomaly, the event is regarded as a false detection. Conversely, when two or more systems detect simultaneously an anomaly, the event is regarded as an anomaly.

UTR-KINKI has two independent and identical safety channels, and each channel uses the UIC in the current mode as a neutron detector. The current signal from the UIC is converted to a voltage signal and displayed as power in percentage on the safety channels #1 and #2 in the control console. The UIC is chosen for the simplest structure and principle among neutron detectors and is expected to have long-term stability and high reliability. Both channels will generate a scram signal to shut down the reactor, when the reactor power exceeds 150% (1.5 W) of the licensed power.