**6. Power monitoring by some developed detectors and new methods**

In this section, several neutron detectors and power monitoring systems are reviewed.

Application of a micro-pocket fission detector for in-core flux measurements is described in section 6.1. SIC neutron monitoring system is examined experimentally and theoretically. Development of an inconel self-powered neutron detector (SPND) for in-core power monitoring will be reviewed in section 6.3. Furthermore, a prototype cubic meter antineutrino detector which is used as a new device for measuring the thermal power as an out-core detection system, will be discussed. Finally, two passive approaches for power measurement are discussed.

### **6.1 Micro-pocket fission detectors (MPFD) for in-core neutron flux monitoring**

There is a need for neutron radiation detectors capable of withstanding intense radiation fields, capable of performing ''in-core'' reactor measurements, capable of pulse mode and current mode operation, capable of discriminating neutron signals from background gamma ray signals, and that are tiny enough to be inserted directly into a nuclear reactor without significantly perturbing the neutron flux. A device that has the above features is the subject of a Nuclear Engineering Research Initiative (NERI) research project, in which miniaturized fission chambers are being developed and deployed in the Kansas State University (K-State) TRIGA Mark-II research reactor (McGregor, 2005). The unique miniaturized neutron detectors are to be used for three specific purposes (1) as reactor power-level monitors, (2) power transient monitors, and (3) real-time monitoring of the thermal and fast neutron flux profiles in the core. The third application has the unique benefit of providing information that, with mathematical inversion techniques, can be used to infer the three-dimensional (3D) distribution of fission neutron production in the core. Micro-pocket fission detectors (MPFD) are capable of performing near-core and in-core reactor power measurements. The basic design utilizes neutron reactive material confined within a micro-sized gas pocket, thus forming a miniature fission chamber. The housing of the chamber is fabricated from inexpensive ceramic materials, the detectors can be placed throughout the core to enable the 3D mapping of the neutron flux profile in ''real-time''. Initial tests have shown these devices to be radiation hard and potentially capable of operating in a neutron fluence exceeding 1019 cm-2 without noticeable degradation. Figure 7 shows a cutaway view of the basic detector concept. It consists of a small ceramic structure, within which is a miniature gas-filled pocket.

string.

Figure 8.

is shown (Blue and Miller, 2008).

Improving the Performance of the Power Monitoring Channel 243

power density profiles can be achieved with as few as five detector triads per detector

Silicon carbide (SiC) is an interesting material for nuclear-reactor power monitor detectors. It has a wide band-gap, small volume and high break down electric field. In addition, SiC is chemically and neutronically inactive. Using SiC power monitors as in-core detectors provides the ability for high counting rate that may help to increase the safety margins of nuclear reactors. To observe the triton response in the SiC p-n diode, a detector with a 1.56 μm LiF converter (with 95% enriched 6LiF) was used. 6Li atoms in the LiF converter may absorb thermal neutrons and generate 2.05 MeV alpha and 2.73 MeV triton particles (6Li(n,3H)ߙ reaction). An 8 μm Al layer was used to minimize damage in the SiC by blocking all alpha particles. However, most tritons have enough energy to pass through this layer and reach the 4.8 μm SiC active layer. The diameter of the LiF converter is 0.508 cm and the SiC diode area is 1.1 mm x 1.1 mm (diode is a square). The active area of the diode is approximately 0.965 mm2. Upon irradiation in the thermal column (TC) facility, one can observe the triton peak in the recorded detector pulse-height spectra and the concomitant triton induced radiation damage on the detector. A schematic of the detector is shown in

Fig. 8. Schematic of side view of SiC detector. The diameter of the LiF converter is 0.508 cm and the SiC diode active area is 0.965 mm2 (diode is a square). Only the active region of SiC

The SiC detector package was connected to a pulse processing system consisting of a preamplifier (ORTEK 142 B) and a digital spectrum analyzer (Canberra DSA 2000). An oscilloscope (Hewlett Packard 54601B, 100 MHz) was used to study the shape of the signal from the amplifier. Bias voltage was provided by the DSA to the detector through the preamplifier. A power monitoring program was used to verify the reactor power that was displayed in the control room. In addition, the degradation of the SiC detectors in the TC's thermal neutron environment was evaluated in terms of dose and dose rate effects. After irradiating the detector at 455 kW, the count rate per kW decreased by a factor of 2 after 11 hr. The I-V characteristics recorded during pre-irradiation and post-irradiation, confirm degradation of the detectors. A theoretical model of the SiC schottky diode detectors was constructed based on MCNP and TRIM computer codes to study the damage induced by tritons for a given diode detector package configuration in the TC's thermal neutron environment. The predicted count rate was compared with the experimental results that were obtained in the TC irradiation field using a charge sensitive preamplifier. The

**6.2 Experimental and computational evaluation of the response of a SiC neutron** 

**monitoring system in a thermal neutron field** 

Fig. 7. Cut away view of MPFD (McGregor, 2005)

A conductive layer is deposited on opposing sides of the device, but not the perimeter. Neutron reactive material, such as 235U, 232Th, 10B, or some material containing 6Li, is applied over the conductive contact(s). Although both sides may be coated with neutron reactive material, only one side needs to be coated for the device to work. The ceramic pieces must be insulators and must not be composed of neutron-absorbing material. For instance, aluminum oxide or oxidized silicon may be used. Connecting wires must be sealed well so that no gas leaks out. Additionally, the ceramic pieces must be sealed with high temperature cement such that the seal integrity is secure within the hostile environment of a reactor core.

By in-core evaluation the device demonstrated excellent count-rate linearity with reactor power. Further, the small size and minute amount of uranium used permitted pulse mode operation without appreciable deadtime distortions or problems. MPFDs have, thus far, shown exceptional radiation hardness to neutrons, gamma rays, and charged-particle reaction products, while showing no performance degradation for devices exposed to neutron fluences exceeding 1019 cm-2. Further, pulse mode operated devices have shown a linear relation to reactor power for neutron fluxes up to 1012 cm-2 s-1, and smaller MPFDs are expected to operate in pulse mode in even higher neutron fluxes. The next generation of MPFDs will be composed of a triad of detectors on a single substrate, one with a 232Th coating, one with a 235U coating, and one with no coating. Such a triad permits monitoring of the fast neutron flux, the thermal neutron flux, and the gamma ray background, all at the same time. Further, the devices behave as point detectors, which greatly simplify data interpretation. Data from such a MPFD array can be converted into a power density map of the reactor core for real-time analysis. Mathematical models are under development that can relate the power density profiles in the reactor's fuel rods to the flux densities at the detector locations. Key to this formulation is the construction of an appropriate response function that gives the flux at any position in the core to the fast neutrons born at an arbitrary axial depth in any of the core fuel rods. Response functions have been derived and used to illustrate the analysis methods. Thus far, modeled results using predicted sensitivities of the MPFDs indicated that the power density in the fuel can be determined provided that appropriate boundary conditions regarding device placement are met. Good matching to

A conductive layer is deposited on opposing sides of the device, but not the perimeter. Neutron reactive material, such as 235U, 232Th, 10B, or some material containing 6Li, is applied over the conductive contact(s). Although both sides may be coated with neutron reactive material, only one side needs to be coated for the device to work. The ceramic pieces must be insulators and must not be composed of neutron-absorbing material. For instance, aluminum oxide or oxidized silicon may be used. Connecting wires must be sealed well so that no gas leaks out. Additionally, the ceramic pieces must be sealed with high temperature cement such that the seal integrity is secure within the hostile environment of a reactor core. By in-core evaluation the device demonstrated excellent count-rate linearity with reactor power. Further, the small size and minute amount of uranium used permitted pulse mode operation without appreciable deadtime distortions or problems. MPFDs have, thus far, shown exceptional radiation hardness to neutrons, gamma rays, and charged-particle reaction products, while showing no performance degradation for devices exposed to neutron fluences exceeding 1019 cm-2. Further, pulse mode operated devices have shown a linear relation to reactor power for neutron fluxes up to 1012 cm-2 s-1, and smaller MPFDs are expected to operate in pulse mode in even higher neutron fluxes. The next generation of MPFDs will be composed of a triad of detectors on a single substrate, one with a 232Th coating, one with a 235U coating, and one with no coating. Such a triad permits monitoring of the fast neutron flux, the thermal neutron flux, and the gamma ray background, all at the same time. Further, the devices behave as point detectors, which greatly simplify data interpretation. Data from such a MPFD array can be converted into a power density map of the reactor core for real-time analysis. Mathematical models are under development that can relate the power density profiles in the reactor's fuel rods to the flux densities at the detector locations. Key to this formulation is the construction of an appropriate response function that gives the flux at any position in the core to the fast neutrons born at an arbitrary axial depth in any of the core fuel rods. Response functions have been derived and used to illustrate the analysis methods. Thus far, modeled results using predicted sensitivities of the MPFDs indicated that the power density in the fuel can be determined provided that appropriate boundary conditions regarding device placement are met. Good matching to

Fig. 7. Cut away view of MPFD (McGregor, 2005)

power density profiles can be achieved with as few as five detector triads per detector string.

#### **6.2 Experimental and computational evaluation of the response of a SiC neutron monitoring system in a thermal neutron field**

Silicon carbide (SiC) is an interesting material for nuclear-reactor power monitor detectors. It has a wide band-gap, small volume and high break down electric field. In addition, SiC is chemically and neutronically inactive. Using SiC power monitors as in-core detectors provides the ability for high counting rate that may help to increase the safety margins of nuclear reactors. To observe the triton response in the SiC p-n diode, a detector with a 1.56 μm LiF converter (with 95% enriched 6LiF) was used. 6Li atoms in the LiF converter may absorb thermal neutrons and generate 2.05 MeV alpha and 2.73 MeV triton particles (6Li(n,3H)ߙ reaction). An 8 μm Al layer was used to minimize damage in the SiC by blocking all alpha particles. However, most tritons have enough energy to pass through this layer and reach the 4.8 μm SiC active layer. The diameter of the LiF converter is 0.508 cm and the SiC diode area is 1.1 mm x 1.1 mm (diode is a square). The active area of the diode is approximately 0.965 mm2. Upon irradiation in the thermal column (TC) facility, one can observe the triton peak in the recorded detector pulse-height spectra and the concomitant triton induced radiation damage on the detector. A schematic of the detector is shown in Figure 8.

Fig. 8. Schematic of side view of SiC detector. The diameter of the LiF converter is 0.508 cm and the SiC diode active area is 0.965 mm2 (diode is a square). Only the active region of SiC is shown (Blue and Miller, 2008).

The SiC detector package was connected to a pulse processing system consisting of a preamplifier (ORTEK 142 B) and a digital spectrum analyzer (Canberra DSA 2000). An oscilloscope (Hewlett Packard 54601B, 100 MHz) was used to study the shape of the signal from the amplifier. Bias voltage was provided by the DSA to the detector through the preamplifier. A power monitoring program was used to verify the reactor power that was displayed in the control room. In addition, the degradation of the SiC detectors in the TC's thermal neutron environment was evaluated in terms of dose and dose rate effects. After irradiating the detector at 455 kW, the count rate per kW decreased by a factor of 2 after 11 hr. The I-V characteristics recorded during pre-irradiation and post-irradiation, confirm degradation of the detectors. A theoretical model of the SiC schottky diode detectors was constructed based on MCNP and TRIM computer codes to study the damage induced by tritons for a given diode detector package configuration in the TC's thermal neutron environment. The predicted count rate was compared with the experimental results that were obtained in the TC irradiation field using a charge sensitive preamplifier. The

Improving the Performance of the Power Monitoring Channel 245

scattered rays. To estimate the gamma field at the detector location, a miniature gamma ion chamber (6 mm diameter and 25 mm long) was used. The calculated gamma sensitivity, 24.8 (fA R-1 h) was used to determine the gamma field at the self-powered neutron detector location. The three SPNDs (inconel600, cobalt, platinum) and were tested together with the miniature gamma chamber in a 200 kCi 60Co source facility. The results showed that the gamma response of the inconel600 and Co detector was found to be similar. However, it was observed that unlike the platinum detector, which has positive response, the Co and inconel detectors showed negative response. The gamma sensitivity of the inconel600 detector is about 7.7 times lower than Pt detector. This low gamma response of the inconel600 detector improves the neutron to gamma ratio and makes it desirable for reactor safety and control applications. In addition to gamma sensitivity, the neutron sensitivity of SPNDs was tested in dry tube (55 mm diameter ×8.4 m long) in-core location of the Pool type reactor. The neutron sensitivity and the total sensitivity of the inconel600 detector were found to be lower than the Co detector. The total sensitivity of the inconel SPD is about 20– 25% of the sensitivity of cobalt and about 35% of the sensitivity of platinum detectors of similar dimensions; however, it is proposed to improve the sensitivity by helically winding the detector with a short axial length. Finally by comparison, the performance of the inconel detector with cobalt and platinum detectors of similar dimensions, it is obvious that inconel

**6.4 Monitoring the thermal power of nuclear reactors with a prototype cubic meter** 

explicitly in terms of the reactor thermal power, Pth. The thermal power is defined as

is the number of fissions per unit time for isotope i, and E�

��� = ∑ ��

released per fission for this isotope. The sum runs over all fissioning isotopes, with 235U, 238U, 239Pu, and 241Pu accounting for more than 99% of all fissions. The antineutrino emission rate ��� (�) can then be expressed in terms of the power fractions and the total thermal power

� � �� �

� (7)

�

is the thermal energy

A new power monitoring method applied to a pressurized water reactors designed by combustion engineering. The method estimate quickly and precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Antineutrino emission in nuclear reactors arises from the beta decay of neutron-rich fragments produced by heavy element fissions, and is thereby linked to the fissile isotope production and consumption processes of interest for reactor safeguards. On average, fission is followed by the production of approximately six antineutrinos. The antineutrinos emerge from the core isotropically, and effectively without attenuation. Over the few MeV energy range within which, reactor antineutrinos are typically detected, the average number of antineutrinos produced per fission is significantly different for the two major fissile elements, 235U and 239Pu. Hence, as the core evolves and the relative mass fractions and fission rates of these two elements change, the measured antineutrino flux in this energy range will also change. It is useful to express the relation between fuel isotopic and the antineutrino count rate

SPD is a useful alternative to Co and Pt SPDs.

**antineutrino detector** 

where N� �

as:

experimental results are in agreement with the predicted response to within a factor of three. I-V measurements show some annealing effects occurring at room temperature. Maintaining the detectors at a higher temperature during irradiation may cause more annealing to occur, thus reducing degradation of the detector. Experiments are necessary to test the degradation of the detector at elevated temperatures, to determine if the effects of annealing are sufficiently great so that the detectors may be useful for neutron power monitoring at high count rates.

#### **6.3 Development of an inconel self powered neutron detector for in-core reactor monitoring**

An inconel600 self-powered neutron detector has been developed and tested for in-core neutron monitoring (Alex, 2007). The sensing material in a self-powered detector is an emitter from which electrons are emitted when exposed to radiation. These electrons penetrate the thin insulation around the emitter and reach the outer sheath without polarising voltage. Some electrons are emitted from the insulator and sheath also. The net flow of electrons from the emitter gives rise to a DC signal in an external circuit between the emitter and sheath, which is proportional to the incident neutron flux. Rh and V SPDs work on the basis of (n, β) reaction and are used for flux mapping while Co and Pt SPDs work on the basis of (n, γ−e) prompt reaction and are used for reactor control and safety. However, the build-up of the 60Co and 61Co gives rise to background signal in the cobalt detector thereby reducing the useful life. In the case of the platinum detector, the detector responds to both reactor neutrons via (n, γ, e) interaction and reactor gamma rays via (γ, e) interaction. Since the neutron sensitivity varies with irradiation as a result of burn up while the gamma sensitivity remains the same, the dynamic response of a mixed response detector varies with time. This mixed and time-dependent response of platinum SPD gives rise to anomalous behaviour in some situations. Development of SPDs with inconel emitters as alternative to Co and Pt prompt SPDs has been reported in literatures. The detector (Figure 9) consists of a 2 mm diameter × 21 cm long inconel 600 emitter wire surrounded by a high purity alumina ceramic tube (2.2 mm ID × 2.8 mm OD). The assembly is enclosed in a 3 mm ID × 3.5 mm OD inconel600 tube.

Fig. 9. Schematic diagram of self powered neutron detector

One end of the emitter is coupled to the conductor of a 2 mm diameter × 12 m long twin core mineral insulated (MI) cable while the detector sheath is laser welded to the MI cable sheath. The detector is integrally coupled to the MI cable and the cold end of the cable is sealed by a twin core ceramic-to-metal seal over which a Lemo connector is fitted.

The gamma sensitivity of the detectors was measured in pure gamma field using 60Co source facility. The detectors were placed at a distance of 1m from the source for better source to detector geometry and 1m above the ground to minimize background from

experimental results are in agreement with the predicted response to within a factor of three. I-V measurements show some annealing effects occurring at room temperature. Maintaining the detectors at a higher temperature during irradiation may cause more annealing to occur, thus reducing degradation of the detector. Experiments are necessary to test the degradation of the detector at elevated temperatures, to determine if the effects of annealing are sufficiently great so that the detectors may be useful for neutron power

**6.3 Development of an inconel self powered neutron detector for in-core reactor** 

An inconel600 self-powered neutron detector has been developed and tested for in-core neutron monitoring (Alex, 2007). The sensing material in a self-powered detector is an emitter from which electrons are emitted when exposed to radiation. These electrons penetrate the thin insulation around the emitter and reach the outer sheath without polarising voltage. Some electrons are emitted from the insulator and sheath also. The net flow of electrons from the emitter gives rise to a DC signal in an external circuit between the emitter and sheath, which is proportional to the incident neutron flux. Rh and V SPDs work on the basis of (n, β) reaction and are used for flux mapping while Co and Pt SPDs work on the basis of (n, γ−e) prompt reaction and are used for reactor control and safety. However, the build-up of the 60Co and 61Co gives rise to background signal in the cobalt detector thereby reducing the useful life. In the case of the platinum detector, the detector responds to both reactor neutrons via (n, γ, e) interaction and reactor gamma rays via (γ, e) interaction. Since the neutron sensitivity varies with irradiation as a result of burn up while the gamma sensitivity remains the same, the dynamic response of a mixed response detector varies with time. This mixed and time-dependent response of platinum SPD gives rise to anomalous behaviour in some situations. Development of SPDs with inconel emitters as alternative to Co and Pt prompt SPDs has been reported in literatures. The detector (Figure 9) consists of a 2 mm diameter × 21 cm long inconel 600 emitter wire surrounded by a high purity alumina ceramic tube (2.2 mm ID × 2.8 mm OD). The assembly is enclosed in a 3 mm

One end of the emitter is coupled to the conductor of a 2 mm diameter × 12 m long twin core mineral insulated (MI) cable while the detector sheath is laser welded to the MI cable sheath. The detector is integrally coupled to the MI cable and the cold end of the cable is

The gamma sensitivity of the detectors was measured in pure gamma field using 60Co source facility. The detectors were placed at a distance of 1m from the source for better source to detector geometry and 1m above the ground to minimize background from

sealed by a twin core ceramic-to-metal seal over which a Lemo connector is fitted.

monitoring at high count rates.

ID × 3.5 mm OD inconel600 tube.

Fig. 9. Schematic diagram of self powered neutron detector

**monitoring** 

scattered rays. To estimate the gamma field at the detector location, a miniature gamma ion chamber (6 mm diameter and 25 mm long) was used. The calculated gamma sensitivity, 24.8 (fA R-1 h) was used to determine the gamma field at the self-powered neutron detector location. The three SPNDs (inconel600, cobalt, platinum) and were tested together with the miniature gamma chamber in a 200 kCi 60Co source facility. The results showed that the gamma response of the inconel600 and Co detector was found to be similar. However, it was observed that unlike the platinum detector, which has positive response, the Co and inconel detectors showed negative response. The gamma sensitivity of the inconel600 detector is about 7.7 times lower than Pt detector. This low gamma response of the inconel600 detector improves the neutron to gamma ratio and makes it desirable for reactor safety and control applications. In addition to gamma sensitivity, the neutron sensitivity of SPNDs was tested in dry tube (55 mm diameter ×8.4 m long) in-core location of the Pool type reactor. The neutron sensitivity and the total sensitivity of the inconel600 detector were found to be lower than the Co detector. The total sensitivity of the inconel SPD is about 20– 25% of the sensitivity of cobalt and about 35% of the sensitivity of platinum detectors of similar dimensions; however, it is proposed to improve the sensitivity by helically winding the detector with a short axial length. Finally by comparison, the performance of the inconel detector with cobalt and platinum detectors of similar dimensions, it is obvious that inconel SPD is a useful alternative to Co and Pt SPDs.

#### **6.4 Monitoring the thermal power of nuclear reactors with a prototype cubic meter antineutrino detector**

A new power monitoring method applied to a pressurized water reactors designed by combustion engineering. The method estimate quickly and precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Antineutrino emission in nuclear reactors arises from the beta decay of neutron-rich fragments produced by heavy element fissions, and is thereby linked to the fissile isotope production and consumption processes of interest for reactor safeguards. On average, fission is followed by the production of approximately six antineutrinos. The antineutrinos emerge from the core isotropically, and effectively without attenuation. Over the few MeV energy range within which, reactor antineutrinos are typically detected, the average number of antineutrinos produced per fission is significantly different for the two major fissile elements, 235U and 239Pu. Hence, as the core evolves and the relative mass fractions and fission rates of these two elements change, the measured antineutrino flux in this energy range will also change. It is useful to express the relation between fuel isotopic and the antineutrino count rate explicitly in terms of the reactor thermal power, Pth. The thermal power is defined as

$$P\_{\rm th} = \sum\_{l} N\_{l}^{f} \, E\_{l}^{f} \tag{7}$$

where N� � is the number of fissions per unit time for isotope i, and E� � is the thermal energy released per fission for this isotope. The sum runs over all fissioning isotopes, with 235U, 238U, 239Pu, and 241Pu accounting for more than 99% of all fissions. The antineutrino emission rate ��� (�) can then be expressed in terms of the power fractions and the total thermal power as:

Improving the Performance of the Power Monitoring Channel 247

days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other

**6.5 Application of Cherenkov radiation and a designed detector for power monitoring**  Cherenkov radiation is a process that could be used as an excess channel for power measurement to enhance redundancy and diversity of a reactor. This is especially easy to establish in a pool type research reactor (the TRR). A simple photo diode array is used in Tehran Research Reactor to measure and display power in parallel with the existing conventional detectors (Arkani and Gharib, 2009). Experimental measurements on this channel showed that a good linearity exists above 100 kW range. The system has been in use for more than a year and has shown reliability and precision. Nevertheless, the system is subject to further modifications, in particular for application to lower power ranges. TRR is originally equipped with four channels, namely, a fission chamber (FC), a compensated ionization chamber (CIC), and two uncompensated ionization chambers (UIC). However, in order to improve the power measuring system, two more channels have also been considered for implementation in recent years. One of these channels is based on 16O (n,p) 16N reaction which is very attractive due to the short half life of 16N (about 7 s). The other channel, at the center of our attention in this work, is based on measurement of Cherenkov radiation produced within and around the core. This channel has a fast response to power change and has been in operation since early 2007. It has been established that the movement of a fast charged particle in a transparent medium results in a characteristic radiation known as Cherenkov radiation. The bulk of radiation seen in and around a nuclear reactor core is mainly due to Beta and Gamma particles either from fission products or directly emanating from the fission process (prompt fission gamma rays). As it will be explained more thoroughly in the following section, Cherenkov radiation is produced through a number of ways when: (a) beta particles emitted by fission products travel with speeds greater than the speed of light in water and (b) indirect ionization by Gamma radiation produces electrons due to photo electric effect, Compton effect and pair production effect. Among these electrons, Compton electrons are the main contributors to Cherenkov radiation. It is established that Cherenkov light is produced by charged particles which pass through a transparent medium faster than the phase velocity of light in that medium. Considering the fact that speed of light in water is 220,000 km/s, the corresponding electron energy that is required to produce Cherenkov light is 0.26 MeV. This is the threshold energy for electrons that are energetic enough to produce Cherenkov light. It is the principal basis of Cherenkov light production in pool type research reactors in which the light is readily visible. For prompt Gamma rays, in general, it makes it possible to assume that Cherenkov light intensity is a linear function of reactor power. It is clear that neutron intensity, fission rate, power density, and total power itself are all inter-related by a linear relationship. In other words, Cherenkov light intensity is also directly proportional to the fission rate. This leads us to the fact that the measured Cherenkov light intensity at any point in a reactor is linearly proportional to the instantaneous power. As long as the measurement point is fixed, the total power could easily be derived from the light intensity with proper calibration. It should be noted here that, as mentioned before, Cherenkov light is also emitted by the electrons produced by the indirect ionization of fission products by Gamma rays, which are confined in fuel elements. For this reason, a linear relationship between reactor power and Cherenkov light intensity would only hold at the higher power range where fission power is dominant in comparison with residual power. Cherenkov light emanating from core is

cooperative monitoring regimes.

$$m\_{\mathfrak{P}}(t) = P\_{th}(t) \sum\_{l} \frac{f\_l(t)}{E\_l^f} \int \varphi\_l \left( E\_{\mathfrak{P}} \right) dE\_{\mathfrak{P}} \tag{8}$$

where the explicit time dependence of the fission fractions and, possibly, the thermal power are noted. �(��� ), is the energy dependent antineutrino number density per MeV and fission for the ith isotope. �(��� ) has been measured and tabulated. Equation 7 defines the burn-up effect. The fission rates �� � (�) and power fractions ��(�) change by several tens of percent throughout a typical reactor cycle as 235U is consumed and 239Pu produced and consumed in the core. These changes directly affect the antineutrino emission rate ��� (�). Reactor antineutrinos are normally detected via the inverse beta decay process on quasi-free protons in hydrogenous scintillator. In this charged current interaction, the antineutrino �� converts the proton into a neutron and a positron: ��� � � �� � �. For this process, the cross section σ is small, with a numerical value of only ~10���cm�. The small cross section can be compensated for with an intense source such as a nuclear reactor. For example, cubic meter scale hydrogenous scintillator detectors, containing ~10�� target protons ��, will register thousands of interactions per day at standoff distances of 10-50 meters from typical commercial nuclear reactors. In a measurement time T, the number of antineutrinos detected via the inverse beta decay process is:

$$N\_{\mathbb{P}}(t) = (\frac{TN\_p}{4\pi D^2})P\_{th}(t)\sum\_{l} \frac{f\_l(t)}{E\_l^f} \int \sigma \,\varphi \rho\_l \,\epsilon \,dE\_{\mathbb{P}} \tag{9}$$

In the above equation, � is the energy dependent cross section for the inverse beta decay interaction, �� is the number of target protons in the active volume of the detector, and *D* is the distance from the detector to the center of the reactor core. � is the intrinsic detection efficiency, which may depend on both energy and time. The antineutrino energy density and the detection efficiency are folded with the cross section �, integrated over all antineutrino energies, and summed over all isotopes i to yield the antineutrino detection rate. The SONGS1 detector consists of three subsystems; a central detector, a passive shield, and a muon veto system. Figure 10 shows a cut away diagram of the SONGS1 detector. Further information can be found in (Bowden, 2007) and (Bernstein et al., 2007).

Fig. 10. A cut away diagram of the SONGS1 detector (showing the major subsystems).

This prototype that is operated at 25 meter standoff from a reactor core, can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7

 ��

where the explicit time dependence of the fission fractions and, possibly, the thermal power are noted. �(��� ), is the energy dependent antineutrino number density per MeV and fission for the ith isotope. �(��� ) has been measured and tabulated. Equation 7 defines the burn-up

throughout a typical reactor cycle as 235U is consumed and 239Pu produced and consumed in the core. These changes directly affect the antineutrino emission rate ��� (�). Reactor antineutrinos are normally detected via the inverse beta decay process on quasi-free protons in hydrogenous scintillator. In this charged current interaction, the antineutrino �� converts the proton into a neutron and a positron: ��� � � �� � �. For this process, the cross section σ is small, with a numerical value of only ~10���cm�. The small cross section can be compensated for with an intense source such as a nuclear reactor. For example, cubic meter scale hydrogenous scintillator detectors, containing ~10�� target protons ��, will register thousands of interactions per day at standoff distances of 10-50 meters from typical commercial nuclear reactors. In a measurement time T, the number of antineutrinos detected

����)���(�) ∑ ��(�)

Further information can be found in (Bowden, 2007) and (Bernstein et al., 2007).

Fig. 10. A cut away diagram of the SONGS1 detector (showing the major subsystems).

This prototype that is operated at 25 meter standoff from a reactor core, can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7

In the above equation, � is the energy dependent cross section for the inverse beta decay interaction, �� is the number of target protons in the active volume of the detector, and *D* is the distance from the detector to the center of the reactor core. � is the intrinsic detection efficiency, which may depend on both energy and time. The antineutrino energy density and the detection efficiency are folded with the cross section �, integrated over all antineutrino energies, and summed over all isotopes i to yield the antineutrino detection rate. The SONGS1 detector consists of three subsystems; a central detector, a passive shield, and a muon veto system. Figure 10 shows a cut away diagram of the SONGS1 detector.

 ��

� � ��� (��� ) ���� (8)

� � � � �� � ���� (9)

(�) and power fractions ��(�) change by several tens of percent

��� (�) = ���(�) ∑ ��(�)

effect. The fission rates ��

via the inverse beta decay process is:

�

��� (�) = ( ���

days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

#### **6.5 Application of Cherenkov radiation and a designed detector for power monitoring**

Cherenkov radiation is a process that could be used as an excess channel for power measurement to enhance redundancy and diversity of a reactor. This is especially easy to establish in a pool type research reactor (the TRR). A simple photo diode array is used in Tehran Research Reactor to measure and display power in parallel with the existing conventional detectors (Arkani and Gharib, 2009). Experimental measurements on this channel showed that a good linearity exists above 100 kW range. The system has been in use for more than a year and has shown reliability and precision. Nevertheless, the system is subject to further modifications, in particular for application to lower power ranges. TRR is originally equipped with four channels, namely, a fission chamber (FC), a compensated ionization chamber (CIC), and two uncompensated ionization chambers (UIC). However, in order to improve the power measuring system, two more channels have also been considered for implementation in recent years. One of these channels is based on 16O (n,p) 16N reaction which is very attractive due to the short half life of 16N (about 7 s). The other channel, at the center of our attention in this work, is based on measurement of Cherenkov radiation produced within and around the core. This channel has a fast response to power change and has been in operation since early 2007. It has been established that the movement of a fast charged particle in a transparent medium results in a characteristic radiation known as Cherenkov radiation. The bulk of radiation seen in and around a nuclear reactor core is mainly due to Beta and Gamma particles either from fission products or directly emanating from the fission process (prompt fission gamma rays). As it will be explained more thoroughly in the following section, Cherenkov radiation is produced through a number of ways when: (a) beta particles emitted by fission products travel with speeds greater than the speed of light in water and (b) indirect ionization by Gamma radiation produces electrons due to photo electric effect, Compton effect and pair production effect. Among these electrons, Compton electrons are the main contributors to Cherenkov radiation. It is established that Cherenkov light is produced by charged particles which pass through a transparent medium faster than the phase velocity of light in that medium. Considering the fact that speed of light in water is 220,000 km/s, the corresponding electron energy that is required to produce Cherenkov light is 0.26 MeV. This is the threshold energy for electrons that are energetic enough to produce Cherenkov light. It is the principal basis of Cherenkov light production in pool type research reactors in which the light is readily visible. For prompt Gamma rays, in general, it makes it possible to assume that Cherenkov light intensity is a linear function of reactor power. It is clear that neutron intensity, fission rate, power density, and total power itself are all inter-related by a linear relationship. In other words, Cherenkov light intensity is also directly proportional to the fission rate. This leads us to the fact that the measured Cherenkov light intensity at any point in a reactor is linearly proportional to the instantaneous power. As long as the measurement point is fixed, the total power could easily be derived from the light intensity with proper calibration. It should be noted here that, as mentioned before, Cherenkov light is also emitted by the electrons produced by the indirect ionization of fission products by Gamma rays, which are confined in fuel elements. For this reason, a linear relationship between reactor power and Cherenkov light intensity would only hold at the higher power range where fission power is dominant in comparison with residual power. Cherenkov light emanating from core is

Improving the Performance of the Power Monitoring Channel 249

value of the present PDA system is valid within ±1% at its nominal power. It is concluded that, at least for the case of research reactors, one can simply increase redundancy and diversity of medium-range reactors by employing the Cherenkov detector as an auxiliary tool for monitoring purposes. It is seen that such a system can provide a stable and reliable tool for the major part of power range, and it can assist in the reactor operation with additional safety interlocks to issue appropriate signals. The advantage of the present detector system over conventional ones is that it is far from the radiation source and thus easily accessible for maintenance and fine tuning. It contains no consumable materials to degrade in long term, and it is relatively inexpensive and simple. Nevertheless, a drawback of the Cherenkov system, which is also true about uncompensated ionization chambers, is

Reactivity is a physical characteristic of the core (based on composition, geometry, temperature, pressure, and the ability of the core to produce fission neutrons) and may be either constant or changing with time. In reactor operation or experiments, signals indicating reactor power (or neutron flux) and reactor period are generally used for direct information on the state of the reactor. However, the most important time dependent parameter is reactivity and continuous information on its value from instant to instant should be highly useful. Since reactivity measurement is one of the challenges of monitoring, control and investigation of a nuclear reactor and is in relation with reactor power measuring. Thus, design and construction of a digital reactivity meter as a continuous monitoring of the reactivity will be reviewed in a research reactor. The device receives amplified output of the fission chamber, which is in *mA* range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in TRR are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. Three experiments for system verification for TRR are; determination of control rod worth, void coefficient experiment, and measuring of reactivity during approach to full power (Khalafi and Mosavi, 2011). For investigating the results of reactivity meter, the reactor power and reactivity plots during the step-wise approach to full power of a particular run of TRR reactor are shown in Figure 13. In this experiment the reactor power was initially stable and critical at 100 kW and a positive reactivity insertion was

**6.6 A digital reactivity meter related to reactor power measuring process** 

its lack of linearity in the low power range.

introduced in the core by changes in control rods positions.

Fig. 13. Power and reactivity plots versus time (Khalafi and Mosavi, 2011).

collected by a collimator right above the core and reflected by a mirror onto a sensitive part of the PDA. Figure 11 shows the integrated system at work, overlooking the core.

Fig. 11. Power measuring channel at work in TRR while receiving Cherenkov light (Arkani and Gharib, 2009).

An important factor to be checked is the system fidelity. This means that the response of the system must be the same when the reactor power is raised or lowered. There is a good fidelity within the linearity range by comparison of the Cherenkov system with the output of CIC power monitoring channel. Moreover, there has been no drift observed in the system in the long run as the system functioned properly for almost 2 years since it was installed. Finally, it is necessary to examine whether the reading from the Cherenkov detector is consistent with other channels. Finally, it is necessary to examine whether the reading from the Cherenkov detector is consistent with other channels. Figure 12 shows its good consistency with other conventional channels (only the fission chamber is shown for the sake of simplicity) within a typical shift operation.

Fig. 12. Comparison of Cherenkov detector output with other regular channels within a typical operation shift of TRR (Arkani and Gharib, 2009).

It is observed that the steadiness and stability of the Cherenkov detector is as good as other existing channels. The 16N counts and pool average temperature are also included as further confirmation of the general behavior of the reactor during the operation. Reasonable stability is observed in the hourly readings of all the channels. Based on statistics, the output

collected by a collimator right above the core and reflected by a mirror onto a sensitive part of

Fig. 11. Power measuring channel at work in TRR while receiving Cherenkov light (Arkani

An important factor to be checked is the system fidelity. This means that the response of the system must be the same when the reactor power is raised or lowered. There is a good fidelity within the linearity range by comparison of the Cherenkov system with the output of CIC power monitoring channel. Moreover, there has been no drift observed in the system in the long run as the system functioned properly for almost 2 years since it was installed. Finally, it is necessary to examine whether the reading from the Cherenkov detector is consistent with other channels. Finally, it is necessary to examine whether the reading from the Cherenkov detector is consistent with other channels. Figure 12 shows its good consistency with other conventional channels (only the fission chamber is shown for the

Fig. 12. Comparison of Cherenkov detector output with other regular channels within a

It is observed that the steadiness and stability of the Cherenkov detector is as good as other existing channels. The 16N counts and pool average temperature are also included as further confirmation of the general behavior of the reactor during the operation. Reasonable stability is observed in the hourly readings of all the channels. Based on statistics, the output

and Gharib, 2009).

sake of simplicity) within a typical shift operation.

typical operation shift of TRR (Arkani and Gharib, 2009).

the PDA. Figure 11 shows the integrated system at work, overlooking the core.

value of the present PDA system is valid within ±1% at its nominal power. It is concluded that, at least for the case of research reactors, one can simply increase redundancy and diversity of medium-range reactors by employing the Cherenkov detector as an auxiliary tool for monitoring purposes. It is seen that such a system can provide a stable and reliable tool for the major part of power range, and it can assist in the reactor operation with additional safety interlocks to issue appropriate signals. The advantage of the present detector system over conventional ones is that it is far from the radiation source and thus easily accessible for maintenance and fine tuning. It contains no consumable materials to degrade in long term, and it is relatively inexpensive and simple. Nevertheless, a drawback of the Cherenkov system, which is also true about uncompensated ionization chambers, is its lack of linearity in the low power range.

#### **6.6 A digital reactivity meter related to reactor power measuring process**

Reactivity is a physical characteristic of the core (based on composition, geometry, temperature, pressure, and the ability of the core to produce fission neutrons) and may be either constant or changing with time. In reactor operation or experiments, signals indicating reactor power (or neutron flux) and reactor period are generally used for direct information on the state of the reactor. However, the most important time dependent parameter is reactivity and continuous information on its value from instant to instant should be highly useful. Since reactivity measurement is one of the challenges of monitoring, control and investigation of a nuclear reactor and is in relation with reactor power measuring. Thus, design and construction of a digital reactivity meter as a continuous monitoring of the reactivity will be reviewed in a research reactor. The device receives amplified output of the fission chamber, which is in *mA* range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in TRR are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. Three experiments for system verification for TRR are; determination of control rod worth, void coefficient experiment, and measuring of reactivity during approach to full power (Khalafi and Mosavi, 2011). For investigating the results of reactivity meter, the reactor power and reactivity plots during the step-wise approach to full power of a particular run of TRR reactor are shown in Figure 13. In this experiment the reactor power was initially stable and critical at 100 kW and a positive reactivity insertion was introduced in the core by changes in control rods positions.

Fig. 13. Power and reactivity plots versus time (Khalafi and Mosavi, 2011).

Improving the Performance of the Power Monitoring Channel 251

providing raw counts to SAND-II computer code, neutron energy spectrum was calculated. The measured and calculated spectrum using neutron detector, MCNP and WIMS codes is

Fig. 14. Axial thermal neutron flux distribution in trap at D6 position (Khalafi and Gharib,

Fig. 15. Detector measured and MCNP and WIMS code calculated neutron spectrum(Khalafi

Spectrum calculations were also checked against measurements. Monte Carlo shows a better prediction while WIMS provides a fair result. It is notable that combination of WIMS/CITATION would be sufficient for neutron flux calculations while Monte Carlo technique should be reserved for the final stages of simulation. A good choice of

shown in Figure 15.

1999)

and Gharib, 1999)

The maximum relative error in three experiments is 13.3%. This error is caused by discrete signal that is transferred to the reactivity meter device. A great portion of the data is lost in the discrete signal and some others in the sampling process. As described in this section, the system of a digital reactivity meter developed on a PIC microcontroller and the personal computer is proved to function satisfactorily in the nuclear research reactor and the utilization of the plant instrument signals makes the system simple and economical. Besides, this device can be used to determine the positive reactivity worth of the fresh fuel and the reflector elements added to the core, effectively. According to the above experiments, the relative error of the digital reactivity meter can be reduced by increasing the sampling frequency of the device. Also by using digital signal processing (DSP) utilities, the rate and accuracy of the reactivity meter can be improved. Because derivative circuits are not used in this device, the error due to the noise that is observed in analog circuits decreases extremely.

### **7. Application of computational codes in simulation, modeling and development of the power monitoring tools**

Some developed codes and simulators for improving the power monitoring will be reviewed in this section. For example, MCNP (monte-carlo n-particle transport code) is developed for neutron detector design, or modeling a fission chamber to optimize its performance

#### **7.1 Computational tools to conduct experimental optimization**

Research reactors need a handy computational tool to predict spatial flux changes and following power distribution due to experimental requirements. Therefore it is important to get accurate and precise information ahead of any modifications. To meet this demand, flux measurements were conducted in case that a typical flux trap inside the core to be allocated. In TRR, one of standard fuel boxes, in position D6 in core configuration of the year 1999, was taken out of the core and a water trap was formed in its place. With the aid of miniature neutron detector (MND) using standard procedure, thermal neutron flux is measured inside the water trap. To calculate the flux and power theoretically, two different computational approaches such as diffusion and Monte Carlo methods were chosen. Combination of cell calculation transport code, WIMS-D5, and three-dimensional core calculation diffusion code, such as CITATION, were used to calculate neutron flux inside the whole core either in two or five energy groups. However, MCNP-4, as a Monte Carlo code, was used to calculate neutron flux again inside the whole core as well as inside the trap (Khalafi and Gharib, 1999). Figure 14 shows axial thermal flux distribution along the D6 position by measurement and computation.

It is obvious from the figure; the both calculation codes are satisfactory and a good agreement exists between detector measurements and code computations. However, diffusion method is a rational choice especially for survey calculation where the Monte Carlo approach is more time demanding. For some consideration, in order to measure spectrum, a fixed point on the midplane along D6 axis was chosen. A variety of foils of different material was selected as measuring windows to determine differential fluxes at specified energy bins. Metal foils such as Ti, Se, Mg, Ni, Al, Co, Au, In, and Fe were selected as energy windows. These foils are sensitive to a part of neutron energy spectrum starting form high energies and ending to thermal energies. Induced activity of each foil is measured based on gamma spectroscopy using high purity Germanium (HPGe) detector. By

The maximum relative error in three experiments is 13.3%. This error is caused by discrete signal that is transferred to the reactivity meter device. A great portion of the data is lost in the discrete signal and some others in the sampling process. As described in this section, the system of a digital reactivity meter developed on a PIC microcontroller and the personal computer is proved to function satisfactorily in the nuclear research reactor and the utilization of the plant instrument signals makes the system simple and economical. Besides, this device can be used to determine the positive reactivity worth of the fresh fuel and the reflector elements added to the core, effectively. According to the above experiments, the relative error of the digital reactivity meter can be reduced by increasing the sampling frequency of the device. Also by using digital signal processing (DSP) utilities, the rate and accuracy of the reactivity meter can be improved. Because derivative circuits are not used in this device, the error due to the noise that is observed in analog circuits decreases extremely.

**7. Application of computational codes in simulation, modeling and** 

**7.1 Computational tools to conduct experimental optimization** 

Some developed codes and simulators for improving the power monitoring will be reviewed in this section. For example, MCNP (monte-carlo n-particle transport code) is developed for neutron detector design, or modeling a fission chamber to optimize its

Research reactors need a handy computational tool to predict spatial flux changes and following power distribution due to experimental requirements. Therefore it is important to get accurate and precise information ahead of any modifications. To meet this demand, flux measurements were conducted in case that a typical flux trap inside the core to be allocated. In TRR, one of standard fuel boxes, in position D6 in core configuration of the year 1999, was taken out of the core and a water trap was formed in its place. With the aid of miniature neutron detector (MND) using standard procedure, thermal neutron flux is measured inside the water trap. To calculate the flux and power theoretically, two different computational approaches such as diffusion and Monte Carlo methods were chosen. Combination of cell calculation transport code, WIMS-D5, and three-dimensional core calculation diffusion code, such as CITATION, were used to calculate neutron flux inside the whole core either in two or five energy groups. However, MCNP-4, as a Monte Carlo code, was used to calculate neutron flux again inside the whole core as well as inside the trap (Khalafi and Gharib, 1999). Figure 14 shows axial thermal flux distribution along the D6 position by measurement and computation. It is obvious from the figure; the both calculation codes are satisfactory and a good agreement exists between detector measurements and code computations. However, diffusion method is a rational choice especially for survey calculation where the Monte Carlo approach is more time demanding. For some consideration, in order to measure spectrum, a fixed point on the midplane along D6 axis was chosen. A variety of foils of different material was selected as measuring windows to determine differential fluxes at specified energy bins. Metal foils such as Ti, Se, Mg, Ni, Al, Co, Au, In, and Fe were selected as energy windows. These foils are sensitive to a part of neutron energy spectrum starting form high energies and ending to thermal energies. Induced activity of each foil is measured based on gamma spectroscopy using high purity Germanium (HPGe) detector. By

**development of the power monitoring tools** 

performance

providing raw counts to SAND-II computer code, neutron energy spectrum was calculated. The measured and calculated spectrum using neutron detector, MCNP and WIMS codes is shown in Figure 15.

Fig. 14. Axial thermal neutron flux distribution in trap at D6 position (Khalafi and Gharib, 1999)

Fig. 15. Detector measured and MCNP and WIMS code calculated neutron spectrum(Khalafi and Gharib, 1999)

Spectrum calculations were also checked against measurements. Monte Carlo shows a better prediction while WIMS provides a fair result. It is notable that combination of WIMS/CITATION would be sufficient for neutron flux calculations while Monte Carlo technique should be reserved for the final stages of simulation. A good choice of

Improving the Performance of the Power Monitoring Channel 253

be used in a low flux neutron surrounding for detection with high resolution. In contrast, by increasing the inter-electrode gap, the fission chamber can be used in a high flux nuclear reactor. Since the pressure variations have significant effects on the sensitivity, the detector components should be designed in accordance with the location, temperature, and neutron flux of the nuclear reactor core. Finally, applying the proper voltage not only enhances the

In addition, the chamber is modeled by GEANT4 to evaluate its sensitivity to gamma ray, which exists as background. Figure 17 illustrates geometry of the modeled chamber in GEANT4. The unwanted noises from gamma ray in the core are dispensable, but in laboratory, this sensitivity must be accounted for the experiments as a disturbance signal.

sensitivity and readout, but also increases the longevity of the chamber.

Fig. 17. Geometry of the modeled chamber in GEANT4

power and the analysis of its behavior.

**8. Thermal methods for power monitoring of nuclear reactor** 

Power monitoring using thermal power produced by reactor core is a method that is used in many reactors. To explain how the method is used for reactor power measurement, a research reactor is studied in this section. In IPR-R1, a TRIGA Mark I Research Reactor, the power is measured by four nuclear channels. The departure channel consists of a fission counter with a pulse amplifier that a logarithmic count rate circuit. The logarithmic channel consists of a compensated ion chamber, whose signal is the input to a logarithmic amplifier, which gives a logarithmic power indication from less than 0.1 W to full power. The linear channel consists of a compensated ion chamber, whose signal is the input to a sensitive amplifier and recorder with a range switch, which gives accurate power information from source level to full power on a linear recorder. The percent channel consists of an uncompensated ion chamber, whose signal is the input to a power level monitor circuit and meter, which is calibrated in percentage of full power. The ionization chamber neutron detector measures the flux of neutrons thermalized in the vicinity of the detector. In the present research, three new processes for reactor power measurement by thermal ways were developed as a result of the experiments. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. The other two methods consist in the steady-state energy balance of the primary and secondary reactor cooling loops. A stainless steel-clad fuel element is instrumented with three thermocouples along its centerline in order to evaluate the reactor thermal hydraulic performance. These processes make it possible on-line or off-line evaluation of the reactor

computational tools would save time a lot in this respect and one is encouraged to perform a comprehensive simulation ahead of design and construction of irradiation facility.

#### **7.2 Optimizing the performance of a neutron detector in the power monitoring channel of TRR**

A fission chamber was utilized for neutron detection in TRR. It was a valuable instrument for in-core/out-core information and the core status monitoring during normal and transient operations. A general theoretical model is presented to calculate the currentvoltage characteristics and associated sensitivity for a fission chamber. The chamber was used in the research nuclear reactor, TRR, and a flux-mapping experiment was performed. The experimental current measurement in certain locations of the reactor was compared with the theoretical model results. The characteristic curves were obtained as a function of fission rate, chamber geometry, and chamber gas pressure. An important part of the calculation was related to the operation of the fission chamber in the ionization zone and the applied voltages affecting two phenomena, recombination and avalanche. In developing the theoretical model, the MCNP code was used to compute the fission rate and the SRIM program for ion-pairs computations. In modeling the source for MCNP, the chamber was placed in a volume surrounded by standard air. Figure 16 illustrates the geometrical details of the MCNP simulation (Hashemi-Tilehnoee and Hadad, 2009).

Fig. 16. The geometrical details of the MCNP simulation. (a) The chamber is placed in a volume surrounded by standard air, (b) the chamber geometrical details, width = 2.5795cm and length = 15.25 cm, (c) the cross section of the fission chamber, (d) the anode, (e) the cathode and (f) the fissile element coating.

The theoretical model together with the mentioned codes was used to evaluate the effects of different applicable variations on the chamber's parameters. An effective approach in decreasing the minimum voltage in the plateau zone, and retaining the chamber in the ionization zone, is to reduce the chamber gas pressure. However, by reducing the pressure, we decrease the gas density. This leads to the reduction of ion-pairs generation rate. Reduction of ion-pairs would affect the sensitivity. At high pressures, the plateau zone width would be extended. This extension needs a stronger electric field, which in turn causes the distortion of the electric field due to space charge effect. Thus, pressure is an important parameter in design considerations. Variations in the enrichment of the fissile element resulted in the enhancement of the fission rate and hence the sensitivity while retaining the applied voltage and plateau zone width. However, surface mass increase would require more applied voltage. Sensitivity of detection of the neutron flux would increase by decreasing the inter-electrode gap. In addition, it increases the width of the plateau zone. This extension optimizes the chamber performance and decreases the detection errors. Furthermore, by decreasing the inter-electrode gap, the fission chamber can

computational tools would save time a lot in this respect and one is encouraged to perform a

**7.2 Optimizing the performance of a neutron detector in the power monitoring channel** 

A fission chamber was utilized for neutron detection in TRR. It was a valuable instrument for in-core/out-core information and the core status monitoring during normal and transient operations. A general theoretical model is presented to calculate the currentvoltage characteristics and associated sensitivity for a fission chamber. The chamber was used in the research nuclear reactor, TRR, and a flux-mapping experiment was performed. The experimental current measurement in certain locations of the reactor was compared with the theoretical model results. The characteristic curves were obtained as a function of fission rate, chamber geometry, and chamber gas pressure. An important part of the calculation was related to the operation of the fission chamber in the ionization zone and the applied voltages affecting two phenomena, recombination and avalanche. In developing the theoretical model, the MCNP code was used to compute the fission rate and the SRIM program for ion-pairs computations. In modeling the source for MCNP, the chamber was placed in a volume surrounded by standard air. Figure 16 illustrates the geometrical details

Fig. 16. The geometrical details of the MCNP simulation. (a) The chamber is placed in a volume surrounded by standard air, (b) the chamber geometrical details, width = 2.5795cm and length = 15.25 cm, (c) the cross section of the fission chamber, (d) the anode, (e) the

The theoretical model together with the mentioned codes was used to evaluate the effects of different applicable variations on the chamber's parameters. An effective approach in decreasing the minimum voltage in the plateau zone, and retaining the chamber in the ionization zone, is to reduce the chamber gas pressure. However, by reducing the pressure, we decrease the gas density. This leads to the reduction of ion-pairs generation rate. Reduction of ion-pairs would affect the sensitivity. At high pressures, the plateau zone width would be extended. This extension needs a stronger electric field, which in turn causes the distortion of the electric field due to space charge effect. Thus, pressure is an important parameter in design considerations. Variations in the enrichment of the fissile element resulted in the enhancement of the fission rate and hence the sensitivity while retaining the applied voltage and plateau zone width. However, surface mass increase would require more applied voltage. Sensitivity of detection of the neutron flux would increase by decreasing the inter-electrode gap. In addition, it increases the width of the plateau zone. This extension optimizes the chamber performance and decreases the detection errors. Furthermore, by decreasing the inter-electrode gap, the fission chamber can

comprehensive simulation ahead of design and construction of irradiation facility.

of the MCNP simulation (Hashemi-Tilehnoee and Hadad, 2009).

cathode and (f) the fissile element coating.

**of TRR** 

be used in a low flux neutron surrounding for detection with high resolution. In contrast, by increasing the inter-electrode gap, the fission chamber can be used in a high flux nuclear reactor. Since the pressure variations have significant effects on the sensitivity, the detector components should be designed in accordance with the location, temperature, and neutron flux of the nuclear reactor core. Finally, applying the proper voltage not only enhances the sensitivity and readout, but also increases the longevity of the chamber.

In addition, the chamber is modeled by GEANT4 to evaluate its sensitivity to gamma ray, which exists as background. Figure 17 illustrates geometry of the modeled chamber in GEANT4. The unwanted noises from gamma ray in the core are dispensable, but in laboratory, this sensitivity must be accounted for the experiments as a disturbance signal.

Fig. 17. Geometry of the modeled chamber in GEANT4
