ATOMIC ENERGY V ^ ^ a L'ENERGIE ATOMIQUE OF CANADA LIMITED V ^ B V DU CANADA LIMITEE

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1 AECL-6403 ATOMIC ENERGY V ^ ^ a L'ENERGIE ATOMIQUE OF CANADA LIMITED V ^ B V DU CANADA LIMITEE THE BASIC CONCEPTS OF A FUEL-POWER DETECTOR FOR NUCLEAR POWER REACTORS Concepts de base d'un detecteur alimente par le combustible dans les reacteurs de puissance G.F. LYNCH Chalk River Nuclear Laboratories Laboratoires nucleates de Chalk River Chalk River, Ontario January 1979 Janvier

2 ATOMIC ENERGY OF CANADA LIMITED THE BASIC CONCEPTS OF A FUEL-POWER DETECTOR FOR NUCLEAR POWER REACTORS by G.F LYNCH Chalk River Nuclear Laboratories Chalk River, Ontario KOJ 1J January AECL-6403

3 Concepts de base d'un detecteur aliments par le combustible dans les reacteurs de puissance par G.F. Lynch Rfisume L'alimentation par le combustible est proposee pour remplacer les flux de neutrons et de rayons gamma employes pour le controle et la securit6 des rgacteurs dans les centrales CANDU. Un detecteur dont la reponse dynamique correspondrait au taux de production de la chaleur dans le combustible est ngcessaire pour repondre aux oesoins de la surveillance de la puissance en coeur. On etudie, dans ce rapport, le concept de 1'adaptation des caracteristiques de reponse d'un detecteur auto-alimente de flux a rriponses mixtes pour qu'elles correspondent aux besoins d'un detecteur id' ' alimente par le combustible. L'Energie Atomique du Canada, Limitee Laboratoires nucleaires de Chalk River Chalk River, Ontario KOJ 1J0 Janvier 1979 AECL-64C3

4 ATOMIC ENERGY OF CANADA LIMITED THE BASIC CONCEPTS OF A FUEL-POWER DETECTOR FOR NUCLEAR POWER REACTORS by G.F. Lynoh ABSTRACT Fuel power is proposed as an alternative to neutron or gammaray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. Chalk River Nuclear Laboratories Chalk River, Ontario KOJ 1JO 1979 January AECL-6403

5 TABLE OF 1 CONTENTS 1. INTRODUCTION 2. CHARACTERIZATION OF SELF-POWERED FLUX DETECTOR RESPONSE 3. FUEL-POWER DETECTOR CONCEPT The Energy Deposited in the Fuel Required Response Characteristics of 4 a Fuel-Power Detector 3.3 Practical Detectors 6 4. CONCLUSIONS 7 5. REFERENCES 9

6 LIST OF TABLES Page TABLE 1 TABLE 2 TABLE 3 Characterization of the Components of a Self-Powered Detector Signal Typical Values for the Energy Deposited in the Fuel of a CANDU Reactor at Equilibrium Burnup Optimum Response Characteristics of a Fuel-Power Detector

7 1. INTRODUCTION The steady increase in nuclear power reactor core size, together with the desire to maximize power output and fuel bunvup, has increased the requirements for in-core power measurements. A recent review article by Bock [1] surveys the practical aspects of in-core monitoring and describes the types of detectors currently in use. In general, because of the relatively high thermal-neutron flux, self-powered detectors are the only practical devices with adequate in-core life for use in CANDU power reactors. with the evolution of the CANDU concept, the use of selfpowered detectors has grown such that over 200 will be used Indeed, in each 600 MWe CANDU-PHW reactor now under construction [2]. Self-powered detectors are mainly used to determine the neutron flux at the point of measurement. These flux readings can be used to produce a three-dimensional flux map, which defines the neutron flux at any point in the reactor core. However, the measurement of neutron flux is not an end in itself, but serves to provide a measure of the local fuel power and hence the heat flux through the fuel sheath. A knowledge of the power being transferred to the coolant is required to optimize the fuel power to avoid the conditions of central fuel melting and/or critical heat flux. The direct measurement of fuel temperature has been suggested as an alternative to self-powered flux detectors for in-core measurements [3]. However, the difficulties associated with instrumenting the relatively large number of fuel bundles in a power reactor would appear to make this approach impractical. Another approach, discussed in this report, would be to tailor the dynamic response characteristics of a selfpowered detector so that its signal matches the heat production rate in the fuel rather than the local neutron or gamma-ray flux.

8 CHARACTERIZATION OF SELF-POWERED FLUX DETECTOR RESPONSE As described in reference [4], a convenient way to characterize the response of self-powered flux detectors in a reactor environment is illustrated in Table 1. TABLE 1 Characterization of the Components of a Self-Pcwered Detector Signal RESPONSE FLUX Prompt Delayed Neutron N n Gamma-Ray r y The parameters N, n, r and y represent the fractions of the total detector signal arising from nejtron-capture (n.y.e) events, neutron activation (n, 3") processes, prompt external gamma-rays, and delayed (mainly fission-product) gamma-rays, respectively. The magnitude of each of these components depends on the detector materials and their geometrical arrangement, the type of rei.tor, and detector location in the core. Over the last decade, considerable effort has been devoted to the development of self-powered detectors which are predominantly sensitive to either the neutron or gamma-ray flux in the reactor core. Of the neutron-sensitive detectors, those with V or Rh emitters have the delayed-neutron component, n, as their dominant response mechanism, and are, therefore, inappropriate for use in fast control and safety systems.

9 - 3 - Another practical the emitter material. neutron flux detector has Co as Although the dominant component at the beginning of life is the prompt-neutron response, N, there is a buildup of the delayed component, n, with irradiation, particularly in the relatively high thermal-neutron flux in a heavy-water moderated power reactor. For example, the buildup of the delayed component limits the useful in-core life of the Co detector to less than five years in a CANDU power reactor. Relatively little success has been achieved so far in the development of predominantly gamma-ray sensitive detectors. Detectors with Zr emitters are commercially available and are typical of this type of device. In general, they suffer from small signal strengths, with the result the neutron-induced signal front tha sheath material contributes a significant component to the total signal. Although detectors with Bi or Pb emitters offer potentially larger that gamma-ray sensitivities, they have only been produced for experimental purposes and problems associated with manufacturing these devices on a commercial high temperature environment have still basis and using them in a to be overcome. Since most of the practical possibilities for singleresponse detectors have been exploited, mixed-response detectors, such as those with Pt emitters, are under active development J5]. With their predominantly prompt response (N+r=0.8), and relatively long useful life in CANDU power-reactor fluxes, Pt detectors have overcome many of the limitations of singleresponse devices. Although the response characteristics of the r. mixed-response detectors are more complex [6], and the signal interpretation in terms of neutron or gamma-ray fluxes is more difficult, the use of Pt detectors in the control and safety systems of CANDU reactors is now well established. With the acceptance of mixed-response detectors, consideration should now be given to signal compositions which

10 - 4 - are optimized for reactor control and safety system applications. The concept developed in this report is that of a self-powered detector whose characteristics are such that its dynamic response matches the power production in the fuel. In the following section, appropriate values for the four parameters, N, n, r and y> are derived for such a device. 3. FUEL-POUER DETECTOR CONCEPT 3.1 The Energy Deposited in the Fuel Table 2 lists typical values for the various components of the energy deposited in the fuel elements of CANDUtype power reactors at equilibrium burnup. The various contributions to the total energy production have been divided into two groups: (i) (ii) those which occur simultaneously with fission (i.e. prompt), and those which result from fission-product ^nd activation - product decay (i.e. de'ayed). From the table it can be seen that approximately 93% of the energy deposited in the fuel is prompt and 7% is delayed, with the delayed component dominated by fission-product decay heat. 3.2 Required Response Characteristics of a Fuel-Power Detector To provide an output signal which is proportional to fuel power, it is necessary to match the detector's response characteristics to the heat-production rate in the reactor fuel. In addition to duplicating the amplitudes of the prompt and delayed components, the delayed component should also have the same dynamic characteristics as the decay heat from the fission products. Thus, the ideal power detector should have a prompt fraction, N + r = 0.93, and a delayed component, y + n = In practice, to match the fission-product decay

11 SOURCE OF ENERGY TABLE 2 Typical Values for the Energy Deposited in the Fuel a CANDU-PHW Reactor at Equilibrium Burnup 17] ENERGY/FISSION (MeV) COMPONENT of CLASSIFICATION GROUP Kinetic Energy of Fission Fragments prompt neutron Kinetic Energy of Fission Neutrons^ prompt neutron Fission Gamma-Rays prompt gamma-ray Neutron Capture Gamma-Rays prompt gamma ray Fission-Product Gamma-Rays delayed gamma-ray Fission-Product g~ Decay delayed e~ decay 6. 8 Neutron Capture P~ Decay delayed &~ decay TOTAL NOTES: There is a delayed neutron component which is insignificant in this context. 2 There is a small additional delayed gamma-ray contribution associated with activation product decay. The actual contribution depends on the specific reactor hardware design.

12 - 6 - heat, the detector's delayed response must be associated with its gamma-ray sensitivity, i.e. Y = 0.07, and the delayed neutron signal must be zero, i.e. n = 0. Although self-powered detectors respond promptly to gamma-rays, the gamma-ray flux in a reactor consists of prompt and delayed components. Boyd and Connor [8] have shown that, at a moderator position in a heavy-water moderated test reactor, the prompt fraction of the gamma-ray flux is 0.67 ±0.01. Thus, for a delayed component y = 0.07, the prompt component of the detector's gamma-ray response will be r = This results in a value for the prompt neutron component of N = Thus, in summary, the ideal fuel-power detector shculd have the response characteristics given in Table 3, i.e. a 79% prompt-neutron sensitivity and a gamma-ray sensitivity of 21%. Optimum Response of a Fuel-Power FLUX Neutron Gamma-Ray TABLE 3 Charact eri sties Detector R E S P Prompt N S E delayed Practical Detectors To achieve the characteristics given in Table 3 requires a judicious selection of detector materials and an appropriate geometrical arrangement. However, in practice, specifying suitable materials which are compatible with the reactor environment and also appropriate for commercial manufacture is not always possible. As it turns out, of the detectors for which irradiation data are available, there ar<* two candidates which have response characteristics that closely match the requirements.

13 - 7 - Self-powered detectors with 95 Mo as the emitter material have been proposed as an alternative to Co detectors, as a prompt-neutron detector, but one which would not have the build-up of a delayed background signal [9]. The neutronsensitive component of this detector would be prompt and, because of the relatively high atomic number of molybdenum, the gamma-ray sensitivity would be positive, unlike that of Co, which is negative [5]. Although there are no experimental results for a detector with a 95 Mo emitter, preliminary data from the irradiation of a natural molybdenum detector have shown that the Mo detector has a gamma-ray fractional response of 0.21 and a prompt-neutron response of The balance of the signal components arise from activation products, and produce delayed effects of a few per cent, which contribute both positive and negative signals. It is anticipated that these background contributions would be much reduced in a detector with an enriched 95 Mo emitter. The second detector whose measured characteristics come close to matching the requirements is a device with a 1.44 mm O.D. Pt-clad emitter [10]. For this detector, the prompt-neutron signal fraction is 0.68 and the gamma-ray component, This results in a detector which is 90% prompt, with the 10% delayed component arising predominantly from gamma-ray processes. Both of the above mentioned detectors have characteristics sufficiently close to those required to indicate that the development of such a fuel-power detector is entirely feasible, once the concept of using a fuel-power monitor in reactor control and/or safety systems becomes accepted. 4. CONCLUSIONS Ti basic concept of a fuel-power detector for nuclear power reactors has been described and the ideal response characteristics for such a device have been derived for a

14 - 8 - CANDU power reactor. Since two existing detectors come close to matching the required characteristics, it is anticipated that such a detector could be developed should the need arise. The next step in the development is to examine the power-reactor control and safety system requirements and confirm that a detector which reflects the heating rate in the fuel has advantages over neutron flux measurements for in-core monitoring of spatial power distribution in reactor cores. Several questions concerning the interpretation of fuel power from measurements made in the moderator include - the effect of boron in the moderator, used to suppress the excess reactivity of fresh fuel; - the change in fuel composition with burnup; - the change in response characteristics with detector burnup, and - the effect of reactivity devices, such as control absorbers, on the local neutron to gamma-ray ratio. It should be noted, however, that these effects influence the signals from all the in-core instruments which are currently in use. Although this report has dealt exclusively with the development of a self-powered detector whose signal is proportional to the power generated in the reactor fuel elements, the basic concept of tailoring the dynamic.response characteristics of a flux detector to meet the in-core monitoring requirements could be extended to other areas. Thus, consideration of the optimum monitoring requirements could have a bearing on future self-powered detector developments.

15 REFERENCES [1] H. Bock, Miniature Detectors for Reactor In-Core Neutron Flux Monitoring, Atomic Energy Review 1_4 ( 1 ), 87 (1976). [2] G.F. Lynch, R.B. Shields, and f.h. Joslin, Environmental Effects on the Response of Self-Powered Flux Detectors in CANDU Reactors, Colloquium paper Cl/02, presented at NUCLEX 75, Basel, Switzerland, October Also issued as Atomic Energy of Canada Limited, Report AECL-5386 (1976). [3] M. Kolb, Fuel Temperature Measurement as an Alternative In-Core Power Measuring Technique, Proceedings of the IAEA Specialists' Meeting on In-Core Instrumentation and Failed Fuel Detection and Location, Mississauga, Ontario, May 1974, Atomic Energy of Canada Limited, Report AECL-5124, (1975] [4] G.F. Lynch, R.B. Shields, and P.G. Coulter, characterization of Platinum Self-Powered D3tectors, IELE Trans. Nucl. Sci. NS 24(1), (1977). Also issued as Atomic Energy o f Canada Limited, Report AECL-5623 (1977). [51 R.B. Shields, A Platinum In-Core Flux Detector, IEEE Trans. Nucl. Sci. NS 20(1), (1973). Also issued as Atomic Energy of Canada Limited, Report AECL-4347 (1973). [6] G.F. Lynch, Soi.ie Theoretical Aspects of Self-Powered Detectors. Proceedings of the IAEA Specialists' Meeting on In-Core Instrumentation and Failed Fuel Detection and Location, Mississauga, Ontario, May 1974, Atomic Energy of Canada Limited, Report AECL-5124, (1975). [71 G.C. Hanna, Section i n Application of Current Technology to 1959 First Reference Design of CANDU, DM-55 (Revised) March 1959 (unpublished AECL information). [8] A.W. Boyd and H.W.J. Connor, Decay of the Gamma-Ray Dose Rate Following a Shutdown in the NRX Reactor, Atomic Energy of Canada Limited, Report AECL-2563 (1966). [9] J.C. Kroon, F.M. Smith, and R.I. Taylor, self-powered Flux Detectors: Status and Prospects, Trans. Am. Nucl. Soc. 23, (1976). ~~ [10] C.J. Allan, Response Characteristics of Self-Powered Flux Detectors in CANDU Reactors, Paper IAEA-SM-226/102, presented at the Second International Symposium on Power Plant Control and Instrumentation, Cannes, France, April Also issued as Atomic Energy of Canada Limited, Report AECL-6171 (1978).

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