Seminar:

Power measurements in fusion reactors

Author: Aljaž Čufar

Mentor: doc. dr. Andrej Trkov, co-mentor: dr. Luka Snoj

Ljubljana 2012

AbstractIn this seminar the main plasma confinement methods are presented and power balance

condition for magnetic confinement plasma is estimated. The most important methods and detectors used for power measurements in today's largest magnetic confinement fusion reactors are described. Some JET diagnostic systems for power measurements are described and since the author and one of the mentors of this seminar collaborate with JET in neutron yield calibration project, the basic principles of this calibration are presented.

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ContentsIntroduction......................................................................................................................................3

Confinement methods.................................................................................................................4Power balance.............................................................................................................................6

Power measurements.......................................................................................................................7Fission chambers.........................................................................................................................8The activation system..................................................................................................................8Plasma profile measurements......................................................................................................9

JET (Joint European Torus)...........................................................................................................10"The Plan".................................................................................................................................11JET diagnostic systems.............................................................................................................11JET neutron diagnostics............................................................................................................12

Calculations to support JET neutron yield calibration...................................................................13Monte Carlo simulations...........................................................................................................13How it is done...........................................................................................................................14Findings so far...........................................................................................................................14

Conclusions....................................................................................................................................15References......................................................................................................................................15

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IntroductionWith technological development and rise of the world's population the demand for energy is

increasing. Today's energy sources should at least in principle satisfy the following conditions: low CO2 emissions, low impact on the environment, high level of safety, reliability and availability independent of weather conditions. Some people suggest that renewable energy sources like wind and solar energy will solve our energy problems and satisfy the above conditions, but according to some calculations [1] these sources of energy fall short of the demands of our growing population. Fusion seems to be a possible long-term source of energy, satisfying practically all of the conditions above, and is therefore a far more probable solution assuming it will eventually work on industrial scale.

Fusion energy is based on fusing light nuclei, such as H or He. For man-made fusion reactors there are three most promising fuel cycles. These fuel cycles are the most important ones because of relatively large cross sections at relatively low energies and are based on fusing of deuterium with tritium (D-T), deuterium with deuterium (D-D) or deuterium with helium-3 nuclei (D-3He). In figure 1 the cross sections for these three fuel cycles are presented.

Figure 1: microscopic cross-sections for fusion of D with three different targets [2]

Fusion reactions for these three fuel cycles are the following:

D+T → He4 +n+17.6 MeV , (1)

D+D → T + p+4.1 MeVHe3 +n+3.2 MeV , (2)

D+ He3 → He4 + p+18.3MeV . (3)The most promising today is the D-T fuel cycle because of the highest cross section at the

lowest energies, which means it is the easiest to achieve. Some of the most important things in commercial power production are abundance, availability, accessibility and reliability of supply of the fuel. Deuterium is relatively abundant as 0.0156% (on atomic basis) of hydrogen is deuterium and hydrogen is widely available since more than 70% of the Earth's surface is covered by water. Another important isotope, tritium, is radioactive with a half-life of 12.3 years therefore it is rare in nature (trace amounts are formed by interactions of atmosphere with cosmic rays [3]) that is why it will have to be produced artificially. Amounts that are artificially produced today are mostly unwanted byproducts of fission power reactors and are not sufficient for powering of the fusion power plants and also fusion power should not rely on fission power reactors. Fortunately tritium

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can be produced from lithium, that is also an abundant element. Production can be done in the blanket of the fusion reactor via neutron capture reactions on Li [4]:

Li6 +n → He4 +T+4.8MeV (4)Li7 +n → He4 +T +n−2.5MeV . (5)

This is called breeding of tritium and it will be done in the breeding blanket with neutrons that are produced by fusion reactions [13]. There is an energy threshold of 2.5 MeV under which reaction (5) is not present while reaction (4) has a relatively high cross section for thermal neutrons. Also 6Li has much higher cross section for reaction with neutrons than 7Li and that is why almost all breeding is done with 6Li nuclei. Reactions (1) and (4) show that all fusion neutrons would have to undergo reaction with 6Li in order to produce sufficient amounts of tritium. Because losses are inevitable neutron multiplication is needed and lead and beryllium can be used for this purpose. Another possible fuel 3He is not radioactive and is also rare on earth but it is thought that it is more abundant on the Moon in the Lunar soil as leftover of the solar wind. But even on the Moon quantities are low. 3He is produced by tritium decay.

All D-T reactions produce neutrons, certain part of D-D reactions produce neutrons and in D-3He plasma there are some D-D reactions part of which again produce neutrons. Neutron production in a fusion reactor is therefore directly proportional to the power production. Because neutrons are practically not affected by electric and magnetic fields they can escape the plasma. Hence they can be easily used for measuring the power of the fusion reactors. The purpose of this seminar is to present the methods of power measurements. In the first part of the seminar I will briefly introduce three concepts of plasma confinement methods in fusion reactors and investigate the power balance of a fusion reactor. In the second part I will describe the system for power measurements at the world's biggest tokamak, JET. In the last part of the seminar I will describe the currently ongoing calibration of JET neutron detectors, in which Slovenian researchers are strongly involved and present investigation of one of the problems being solved within this task.

Confinement methodsBecause the cross sections for fusion reactions are quite small and because nuclei

electrically repel each other it is crucial that high enough densities and kinetic energies of nuclei are obtained so that fusion of the nuclei can take place. High energy of the particles means high temperature and temperatures of fusion fuel are so high that the fuel is in the form of plasma. To obtain sufficient densities plasma must be confined and confinement also reduces the necessary heating power by reducing losses of the energetic particles. Because of the high temperatures plasma can not be simply confined by just putting it in the chamber because the plasma facing material would melt immediately. In addition heat transfer from plasma to wall would cause high energy losses and also contamination of plasma with the plasma facing material. There are two main methods of the fusion plasma confinement that can be used: magnetic confinement and inertial confinement. The main focus in this seminar will be on the magnetic confinement devices especially tokamak type since it looks most promising to be utilized in a commercial fusion power reactor.

Tokamak is a kind of toroidal magnetic confinement device. It was invented in 1950's by Soviet physicists Igor Tamm and Andrei Sakharov inspired by idea of Oleg Levrentyev. It has strong toroidal and poloidal magnetic fields so the resulting field is helically twisted around the torus. Toroidal component of the magnetic field is generated by the electromagnets that are positioned all around the torus and the poloidal component of the magnetic field is generated by the electric current through the plasma that is generated by the transformer with the magnetic induction.

Stellarator is another kind of toroidal magnetic confinement device that was invented in

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1950 by an American scientist Lyman Spitzer. It has evolved as an attempt to confine plasma with strong toroidal magnetic fields without electric currents through the plasma. But since purely toroidal field cannot provide the balancing force against the expansion of the plasma, the magnetic field is twisted as it passes around the torus. One of the disadvantages of the stellarators in comparison with the tokamaks is that their shape is more complex with fewer symmetries and because of that the production of the components and the calculations of the particle transport are more complicated.

Figure 2: schematic view of tokamak [4] Figure 3: schematic view of stellarator [4]

Direct/Indirect drive inertial confinement device is a fusion reactor that can use lasers, beams of electrons or beams of ions to ignite the fuel. The most advanced reactor of this type is NIF (National Ignition Facility located at the Lawrence Livermore National Laboratory in California, USA) that uses lasers. The fuel in this reactor is a mixture of solid D and T in a shape of small (millimeters) spherical pellets.

Figure 4: schematic of one of the laser amplifiers of NIF [4]

The powerful laser pulse heats the fuel pellet from all sides (it is very important that the surface is heated as uniformly as possible) the material on the surface explodes and this explosion causes the rest of the fuel to implode. Also a shock wave that travels toward the center of the pellet is created and when it reaches the center the central part of the pellet becomes dense and hot enough for fusion to occur [10]. This is the principle of a direct drive inertial confinement fusion reactor. To

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achieve the needed uniformity of heating the fuel pellet can be put inside a hohlraum (German word for cavity or hollow area). Lasers are pointed at the interior of the hohlraum which absorbs energy of laser beams and re-emits this energy as x-rays that are much more uniformly distributed on the surface of the fuel pellet than direct laser light would be and this means that the explosion and the implosion that are described above are more uniform and more of the fuel undergoes fusion resulting in better efficiency. Some efficiency is lost due to the fact that fuel is heated indirectly. This is a principle of the indirect drive inertial confinement reactor. NIF was primarily designed as an indirect drive device but because it's lasers are powerful enough it can also be used as a direct drive device [10].

Power balanceIn tokamaks it is important to quantify power balance conditions in the fuel, as is the case

with all heat sources. One of the most important parameters is confinement time that measures the rate of energy losses of the system. It is defined as

τE=WP L

(6)

where W is the energy content of the system and P L is the power loss of the system.

Out of 17.6 MeV that is released in a D-T fusion reaction 14.1 MeV is carried away by neutron and 3.5 MeV by α particle. Since neutrons escape out of the magnetic confinement and contribute to wall heating (in the fusion power reactors practically all of the neutrons will be absorbed in the reactor wall and heat deposited by these neutrons will be used for production of the electricity) only α particles provide heating of the plasma and power balance requirement states that the power loss - P L has to be balanced by the fusion α particle heating - Pα and external heating - P H :

P L=Pα+P H . (7)Using equation (6) P L can be calculated because we know that the average energy of the particle is 3

2 kT or 12 kT per degree of freedom (low kinetic energies of the electrons and ions compared to

the rest energies of these particles allow us to use the classical statistical mechanics) and because there is an equal number of electrons and ions the total energy in the plasma is

W =∫3nkT dV=3 nkT V . (8)In the fusion community temperature is usually measured in eV which means that kT is what we have in mind when we use T as temperature and using this convention and equation (8) for energy of the plasma power loss can be written as

P L=WτE

=3 nTτE

V . (9)Pα can be calculated from the equation for power density in the reactor

p=nD nT ⟨σ v⟩ E DT (10)where nD stands for deuterium number density, nT tritium number density, E DT represents D-T fusion energy release and ⟨σ v⟩ is the product of fusion cross-section and relative velocity of deuterium and tritium nuclei averaged over all energies and volume of the reactor. nD nT ⟨σ v⟩ is actually the fusion reaction rate.

Pα=∫ nD nT ⟨σ v ⟩ Eα dV (11)where Eα represents energy of α particles. For plasma with equal deuterium and tritium densities where nD=nT=

n2 the total α particle heating is

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Pα=14

n2 ⟨σ v⟩ Eα V . (12)

Equation (7) can now be written as

P H=( 3nTτE

− 14

n2 ⟨σ v⟩ Eα)V . (13)

Since ignition of plasma means that the plasma burn is self sustaining, this is adopted as the definition of the condition for the ignition. If we simplify the equation by taking plasma density and temperature as constants we get the condition for the ignition as

τE n< 12⟨σ v ⟩

TEα

. (14)

It turns out that in the kT range from 10 to 20 keV ⟨σ v⟩ can be relatively accurately represented by the equation

⟨σ v⟩=1.1×10−24 T 2 m3 s−1 , (15)where T is in keV. Using this equation the ignition condition can be represented by the triple product:

n T τE>3×1021 keV s m−3 . (16)

The triple product is actually the most important parameter when it comes to magnetic confinement fusion. As seen from equations above it's concept is valid only for temperatures in the range between 10 and 20keV.

Another important fusion plasma parameter is the fusion power gain Q that is defined as

Q=P produced

Pused, (17)

where P stands for power. At the moment the best obtained Q was around 0.65 at the JET tokamak [17] and for ITER it is expected that it will reach around 10 [16]. To measure Q it is necessary to have reliable measurements of the produced fusion power. For NIF it is stated that "fusion targets will release 10 to 100 times more energy than the amount of laser energy required to initiate the fusion reaction" [8] But NIF is even further away from a power plant than ITER will be because NIF laser efficiencies are less than 1% (besides other losses) and for a sufficient power production the lasers would have to produce around 10 pulses per second.

Power measurementsSince D-D and D-T fusion reactions release neutrons, reactor physics and neutron transport

are very important. Especially due to the fact that fusion reactors produce more neutrons per unit power than fission reactors. In a fission reactor on average 2.4 neutrons are released per fission reaction releasing approximately 200 MeV while in a fusion reactor every fusion reaction releases a single neutron and 17.6 or 3.25 MeV for D-T or D-D fusion respectively. Moreover fusion neutrons are born with higher energies on average, hence the neutron spectrum in a fusion reactor is harder and thus more damaging for the structural components of the reactor. Also in fusion reactors basically all neutrons hit the first wall because plasma in the reactors is almost a void for neutrons while in fission reactors neutrons drive the chain reaction so a great part of the released neutrons is absorbed in the fuel and the part that actually hits the reactor wall and structure materials is greatly slowed down, since the fuel is surrounded by a moderator. Because one neutron is produced per fusion reaction and the released energy is known the released energy is easily calculated if the rate of neutron production is known. This is why power measurements are based on neutron flux measurements [6]. There are also other methods of power measurements but these methods are

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calibrated on neutron based power measurements. Since all D-T and certain part of D-D fusion reactions produce neutrons which are neutral particles, therefore the electric and magnetic fields do not change their trajectories, it is quite intuitive that the neutron flux measured just outside the reactor is linearly proportional to the fusion power produced in the reactor. Because of the harsh conditions inside the reactor neutron detectors would not last long enough if put inside the reactor. Losses of neutron flux in the reactor blanket are inevitable but also scattering from some large objects like transformer limbs can cause the increase of flux in some areas. That is why it is important to recalibrate detectors after big changes in the reactor structure.

Fission chambersThe most important neutron and main power detectors in magnetic confinement fusion

reactors are fission chambers. Fission chambers are usually similar in construction to ionization chambers. For neutron detection ionization chambers filled with BF3 or 3He gas can be used. 3He is not used as often as BF3 because of the limited supply. Boron lined proportional detectors are also used. The main difference and the reason for greater sensitivity of fission chambers to neutrons when compared to BF3 filled or boron lined ionization chambers is that instead of boron, highly enriched 235U or 238U is used because the fission fragments cause many more ionizations per interaction in the chamber than an alpha particle from boron [12]. Because of a far greater response to neutrons that cause fission in the chamber than to gamma rays fission chambers can be used in higher gamma fields than uncompensated ion chambers with boron linings and that makes fission chambers very appropriate for reactor use.

To calculate fusion power from measured neutron flux, calibration factors are used. Calibration factors are obtained by calibrations during which responses from calibrated neutron sources positioned around the reactor are measured.

The activation systemThe basis for this technique is the phenomenon that materials irradiated with neutrons

commonly become radioactive (they get activated) and then release excess energy by emitting characteristic gamma rays, which can be detected with semiconducting (Ge) detectors. This system consists of samples of various materials (e.g. In) that are positioned near the plasma during pulses and are retrieved for analysis shortly after the pulse. Materials used for samples are contained in capsules and have known characteristics (dimension, mass, material composition) that allow us to determine the neutron fluence end energy spectrum to which the sample has been exposed. Two good characteristics of activation detectors are the large dynamic range of the measurements and the lack of mechanical and electrical components that are sensitive to high yields of radiation. Unfortunately the sensitivity is relatively low and the lowest fluences that can be detected are in the order of 106 - 107cm-2 [9]. Error of the neutron yield measurement in large tokamaks using activation methods is usually smaller than 10%. Total neutron production, or total fusion power is calculated by calibration factors that are determined by neutron transport calculations using Monte Carlo neutron transport codes.

For the analysis of irradiated (activated) materials scintillation or semiconducting detectors are used. The saturation activity of the irradiated samples A0 is given by the following expression:

A0=N p/t m

SDC1

ϵp C COI Pemission, (18)

where N p is the peak area of a particular gamma ray peak in the recorded gamma spectrum of the sample, tm the measurement time, ϵp the peak detection efficiency of the detector, CCOI the coincidence correction factor and Pemission the gamma emission probability. The saturation factor S

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is defined as:

S=1−e−λt r , (19)where t r is the irradiation time and λ is the decay constant of the considered nuclei (typically indium-115 is used as a target to produce the metastable indium-115m nuclei with half life of 4.5h and the decay constant λ= ln2

4.5h ≈0.15/h by the inelastic scattering of neutrons [9]). The decay factor takes into account that measurement is conducted t cool after the irradiation

D=e−λt cool . (20)The fact that nuclei decay during the measurement is taken into account by the counting factor

C=1−e−λ t m

λ tm. (21)

From A0 the reaction rate (R.R.) can be calculated by the equation:

R.R.=Φ0∫0

∞

ϕ(E )σ(E )dE=A0

N a, (22)

from reaction rate the total neutron flux can be calculated:

Φ0=A0

N a∫0

∞

ϕ(E )σ(E )dE , (23)

where ϕ(E ) is the neutron spectrum normalized to the unity integral, Φ0 is total neutron flux,σ(E ) is the absorption cross-section and N a is the number of the target nuclei in the sample [9]. Neutron yield is then calculated from the measured total flux using calibration factors similarly as for the fission chambers.

Alternative methods that would allow using activation system for monitoring neutrons are being developed. One option is to use the reactor cooling water as an activation medium. The water is activated by neutrons via 16O(n,p)16N reaction. The Cherenkov light from beta decay of 16N which have a half life of 7.13s would then be collected near the reactor and transmitted by an optic fiber to the detector that would be positioned outside of the high flux region of the reactor [11].

Plasma profile measurements

Figure 5: neutron profile monitor [7]

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In addition to power, which is an integral parameter of the reactor, one is often interested in the spatial distribution of fusion neutrons. Plasma profile measurements are used for these purposes. Since the plasma cross-section does not vary much around the tokamak a single cross-section of plasma is enough to learn about it's properties, such as power density, ion and electron temperatures, etc. The ports of the reactor ("windows" into plasma) vessel are used for measurements of the plasma profiles. With enough detectors on the right positions 2D plasma cross section properties can be calculated using the same methods that are used in X-ray computer tomography in medicine.

Neutron profile monitor is a system of several neutron detectors and collimators that measure neutron flux from a certain part of plasma. This way fusion power profiles can be obtained [7].

Bolometry uses similar principles as neutron profile monitors. Bolometers are simple detectors that measure total radiation losses of the plasma. They are basically just small pieces of metal with precisely defined thermal properties. These pieces of metal are heated up due to plasma radiation that comes through pinholes (narrow slits) and the increase of the bolometer temperature tells us how great the plasma losses are.

Today's most advanced fusion plasma diagnostic system is used by JET - Europe's largest fusion device.

JET (Joint European Torus)

Figure 6: computer generated model of JET [5]

JET is currently the largest tokamak in the world and the only magnetic confinement device that is capable of operating with D-T fuel [5]. JET started operating in 1983 and was designed to study plasma behavior in conditions close to those required for fusion reactors and to be flexible, so it has been evolving over the years and is currently still state-of-the-art - JET is holding the world record in produced fusion power (16MW during 1997 D-T experiments). Now that ITER is being built JET's primary task is to act as a test bed for the construction and operation of ITER. Even though JET can use D-T fuel, most of the experiments are done with D-D fusion. Reasons for that are the following: neutrons from D-T fusion have higher energies than D-D neutrons - neutrons with lower energies are easier to shield from because they cause less damage and activation in the

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material and that can prolong the period after which the most exposed parts of the reactor have to be replaced, in D-T cycle there is more T which is radioactive gas, so it requires special handling, and D-T reaction at given conditions produces more power than D-D because of the higher cross-section and because of higher energy gain but more produced power means more cooling is needed and more cooling means less space for diagnostics.

JET is a large device of approximately 15 meters in diameter and 12 meters high. It's vacuum vessel has the major radius of 2.96m and the minor radii of 2.10m in vertical and 1.25m in horizontal direction [7]. Cross-section of the vacuum vessel is D-shaped (2.5m × 4.2m). It's additional heating capabilities are 30MW and can produce flat-top pulses that last 20-60s. Toroidal magnetic field of 3.45T is generated by 32 large D-shaped coils with copper windings that consume up to 380MW of electricity. There is also a large transformer that produces poloidal field and other large consumers of electricity, so the total installed capacity of power supply is 1400MW. Since it is not practical to get all this power directly from the grid (this is in the order of magnitude of a large nuclear power plant and you can't just turn on and off nuclear power plant whenever you want) two large flywheels are there to store the energy in order to meet large power requirements during the pulses without causing too great peaks in power requirement from the electricity grid.

"The Plan"One of JET's missions was to prove the feasibility of fusion as a power source. It has shown

that fusion power of the same order of magnitude as the power needed for heating plasma can be generated. Next step is ITER that will try to show that fusion can be controlled and used on the industrial scale and that we can use fusion to generate more power that is required for maintaining of the plasma. If there is a significant progress in understanding and controlling of the plasma, ITER might even be able to cross the ignition point which is the point where heating from alpha particles that are generated with fusion reactions provides enough heating power and no external heating is required. Now ITER is already being built and the first plasma is expected in 2019. ITER will be followed by DEMO, demonstration power plant, that will demonstrate production of electricity using fusion power as a power source.

JET diagnostic systemsApart from neutron measurements JET has many other plasma diagnostic systems. JET's

diagnostic systems and it's experience is very important for ITER's diagnostic systems, since JET comes the closest to ITER's conditions. JET is an experimental reactor and because of that it has many more diagnostic systems that commercial power reactors will have and these systems are constantly being changed, upgraded and replaced. It's all thanks to it's flexible design. Some of the main tasks for JET's diagnostic system are detection of fusion products (neutrons, gamma rays and alpha particles), measurement of plasma isotopic composition, which is important because all hydrogen isotopes can be used, determination of plasma parameters such as temperature, density, particle losses, radiation losses, measurements of magnetic topology and observations of plasma flows and fluctuations [7]. Data from the diagnostic systems is also used for real time control of plasma. This is an important feature that became available with progress in computer technology. Before powerful enough computers real time control was "hard wired" in the system using some simple electronic and results of measurements were shown on oscilloscopes and on photographs. Nowadays all the signals from diagnostic systems are digitized and stored in the central database. During one JET pulse up to 18GB of raw diagnostic data can be produced by around 100 individual instruments that form the diagnostics system [5]. Both active and passive diagnostic methods are used.

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Figure 7: overview of JET diagnostics [7]

JET neutron diagnosticsJET has 3 pairs of moderated ion chambers containing 235U and 238U located in moderator

packages by the transformer magnet limbs that can operate in pulse counting and current modes. Pulse counting is used for lower flux measurements where individual pulses can be separated and a current mode is used for high flux measurements where pulses can't be separated anymore and instead of that the current is measured. This system can detect neutron emission rates in range 1010 - 1018 n/s and chambers that contain 235U are insensitive to neutron energy [6]. Fusion power is calculated from detector response using calibration factors that are obtained during calibrations.

JET's activation system pneumatically delivers samples that are housed in polyethylene capsules [9] into irradiation locations, the so called "Irradiation Ends". After the pulse pneumatic system retrieves the sample so that data about the neutron fluence (integral of flux over time) and spectrum of the pulse is collected with gamma spectroscopy or by counting of delayed neutrons, the method depends on the type of the sample. Sometimes materials that produce beta decaying nuclei when irradiated by neutrons are used. There are 8 irradiation ends in 5 octants of the tokamak [6]. 7 irradiation ends are located behind the tokamak shield and one is located inside the vacuum vessel [9]. The irradiation end that is positioned inside the vacuum vessel is cooled with water cooling system because the temperatures there during pulses are around 300°C. For a material to be useful as an irradiation sample it is important to have high enough cross-section for reaction with neutrons in suitable energy range and at the same time activation products must have appropriate half-lives and must emit gamma photons that are suitable for detection.

Neutron profile monitor in JET consists of two cameras allowing observation of plasma from ten horizontal and nine vertical positions [7].

ITER's diagnostic system will be based on experience from JET but because fusion power in ITER will be much higher and higher fusion power means higher neutron flux thick shielding of superconducting magnets and surroundings of reactor will be necessary. While main JET's neutron monitors are positioned outside the reactor main ITER neutron monitors will not be fission chambers positioned outside the vacuum vessel but in-vessel micro-fission chambers most likely

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installed in a gap between shielding blanket and vacuum vessel [15] [16].

Figure 8: schematic view of a single micro-fission chamber with a dummy chamber for the gamma-ray effect compensation [15]

Calculations to support JET neutron yield calibrationLast calibration of JET's fission chambers was performed in 1980's and has been maintained

over the years by cross-calibration to the in-vessel activation system [17]. Because there were some major changes in the reactor namely transition of the "first-wall" material from carbon to ITER-like material during the 2010-2011 shut-down recalibration of neutron detectors is planned. It is a project in which neutron yield detectors will be calibrated using 252Cf calibrated spontaneous fission neutron source. Scientists from Jožef Stefan Institute collaborate on this project and I am also working on it. The confirmation of the neutron yield calibration will be ensured by direct measurements where Cf neutron source will be deployed on more than 200 positions in the reactor vessel using remote handling manipulator and the detector response will be measured for few minutes on each source position. Since the remote handling system (RH) consists of substantial objects, it is expected that it will significantly affect the measured neutron yields. The specific objective of the present work is to calculate the effects of the JET RH system on the neutron monitors. We developed a simplified geometrical computational model of the JET RH system using the Monte Carlo neutron transport code MCNP.

Monte Carlo simulationsMonte Carlo transport code is widely used in neutron transport calculations. It can simulate

a wide variety of geometries with many different types of sources and materials using neutrons with a wide spectral range. Different cross-section data libraries can be used for solving the same problem and that allows us to see how the results are affected by small changes in cross sections that are used for calculations. Besides neutron transport and interaction electron and photon interactions can also be a part of the program.

In contrast to deterministic methods that give us data about behavior of average particles, Monte Carlo methods actually simulate individual particles and then the program provides us with statistical data about the particles simulated (in the input we define what statistical data we want in the output of the program). Monte Carlo methods are widely used in calculations of the problems that are too complex to be calculated using deterministic methods. Neutron transport calculations in JET with it's ports, massive objects like transformer limbs and remote handling equipment positioned inside the reactor are good examples of a problem in which reliable deterministic calculations would be very difficult.

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How it is doneA simplified Monte Carlo transport code model of the remote handling (RH) was developed.

This model of tokamak was developed to be accurate enough to study all the main effects of the ports and massive objects while at the same time it is simplified enough that the calculation time is reasonably low. All components that have major influence on the neutron yield measurements such as ports and transformer limbs are included in simplified form. The specific objective of the present project is to calculate the effects of the JET RH system on the neutron detectors (especially fission chambers located outside the tokamak) as the ratio of the neutron fluxes calculated with and without the remote handling system present in the model.

When measured neutron detector responses are obtained, it will be possible to calculate what responses would be measured without the remote handling in the reactor by this equation:

ϕ0=ϕRH

C RH, (24)

where ϕRH is the measured detector response and ϕ0 is the response that would be measured if there was no remote handling in the reactor. This way it will be known what are the responses to calibrated neutron source at certain locations inside the reactor. From this approximation of the response to a point source on many locations in the empty reactor the detector response to continuous neutron source like fusion plasma can be calculated.

Figure 9: a simplified model of the tokamak with remote handling [14]

Findings so farA study has been made, which has showed that only 5-10% of the neutrons that hit the

fission chambers come directly through the reactor wall, the other neutrons come through the ports. The biggest contribution to neutron flux comes through the port closest to the fission chamber and the second most important part of the flux comes through the port that is closest to the neutron source (neutrons scatter on a concrete wall that is built all around the tokamak to shield the staff and environment from radiation). Effects of large objects are being studied and some preliminary results about the influence of remote handling on measured neutron yield have been calculated. These preliminary correction factors help us understand the main effects on the neutron transport through the reactor and help us determine how detailed must a model of a certain component of the reactor be, which position of movable massive objects causes the least interference, on which neutron source positions should RH position be changed in order to cause less interference.

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ConclusionsAs one fusion reaction produces one neutron, which can escape the plasma, neutron

measurements are the basis for determination of the fusion reactor power. In order to increase the accuracy of the power measurements, various neutron detectors are used. In addition all measurements are strongly supported by Monte Carlo neutron transport calculations which help in understanding the detector response, calculations of potential correction factors and evaluation of uncertainties. The currently on-going project of fusion yield calibration at JET will provide guidelines and become a reference on how to perform fusion power calibrations of future fusion reactors, namely ITER and DEMO.

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