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Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO
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NUCLEAR ENERGY Annals of Nuclear Energy 33 (2006) 1134–1140 www.elsevier.com/locate/anucene
Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO Myong-Seop Kim *, Byung-Chul Lee, Sang-Jun Park, Byung-Jin Jun Korea Atomic Energy Research Institute, 150, Deokjin-Dong, Yuseong, Daejeon 305-353, Republic of Korea Received 10 March 2006; received in revised form 23 June 2006; accepted 23 June 2006 Available online 28 August 2006
Abstract The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF3 detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF3 detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 · 105 fissions/ cm3 s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 lg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 lg-U235/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time. 2006 Elsevier Ltd. All rights reserved.
1. Introduction In research reactors as well as nuclear power plants, any fuel failure that causes a release of fission products into the primary coolant should be detected as soon as possible. In HANARO, 30 MW research reactor, the fuel failure had been monitored using the gamma-ray detection system. Three NaI(Tl) detectors were placed under the common return line of the primary cooling circuit in order to monitor the gamma-ray activity in the coolant caused by the release of the fission products. From the operation experiences and various analyses, it was ascertained that the fuel failure detection system
*
Corresponding author. Tel.: +82 42 868 8666; fax: +82 42 868 8341. E-mail address: [email protected] (M.S. Kim).
0306-4549/$ - see front matter 2006 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2006.06.002
(FFDS) using NaI(Tl) detector was facing difficulties in monitoring the fuel failure because of N-16, a major gamma-ray source in the coolant. The energies of the gamma-rays from the fission and activation products except N-16 are mostly smaller than 2 MeV. Measuring the total gamma-ray activity below 2 MeV was not sensitive enough to detect the fuel failure due to the high Compton background to the detector of the high energy gamma-rays from the N-16. If a decay system is installed to decrease the N-16 effect, the detection of fuel failure would be delayed in time. Therefore, in HANARO, the N-16 effect had been compensated for by using the predetermined Compton background and the ratio of the net count rate of the gamma-rays with 0.6–2.0 MeV to that above 2.0 MeV. This method provided the sensitive detection of the fuel failure, but the net count rate was not so stable because of the statistical fluctuation. Also, the ratio
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
2. Neutron measurements and verification of their origins In order to measure the neutrons around the primary cooling circuit of HANARO, BF3 proportional detectors were used. The effect of the high energy gamma-rays on the neutron pulse spectrum of the BF3 detector was investigated. The diameter of the BF3 detector is 25 mm, and its sensitive axial length is 120 mm. Fig. 1 shows the pulse height spectrum from the BF3 detector installed below the coolant outlet pipe. Since the neutron pulses are well discriminated from the gamma-ray pulses and the electronic noises, the effect of high energy gamma-rays on the neutron pulse spectrum of the BF3 detector is negligible. Therefore, the BF3 detector could be installed at the position as close as possible to the coolant outlet from the core, i.e., the delaying in time for detection of fuel failure was minimized. The time sensitivity of this system was improved much more than that of previous FFDS using NaI(Tl) detector that should be installed around the pipe of the return line in
12000 10000
Counts/channel
was rapidly increased when the reactor power was being decreased because the decay of the activation products was slower than that of the N-16, which caused an unscheduled reactor trip. Another possibility to monitor the fuel failure is the detection of the delayed neutrons from the short-lived fission products such as Br-87 and I-137 (Dobrin et al., 1997; Jacobi et al., 1977). These delayed neutron precursors are released to the coolant by a recoil process from the inside of the failed fuel rods (Venkateswaran and Venkateswarlu, 1989). During normal reactor operation, the neutrons can be generated in and out of primary cooling circuit from various sources such as streaming through the coolant duct, photodisintegration by high energy gamma-rays from activated nuclei like N-16 and Na-24, and delayed neutron generation from contaminated uranium on fuel surface (Jacobi et al., 1977; Kawai et al., 1990; Venkateswaran and Venkateswarlu, 1989). In order to apply the FFDS based on the delayed neutron measurement to HANARO operation, it should be confirmed that the rapid and sensitive detection of neutron from the inside of the failed fuel rod is possible regardless of the background contributions of above neutrons. Therefore, in this work, the neutrons from the primary coolant during the normal operation of HANARO were measured, and the origins of these neutrons were analyzed. Then, the detection sensitivity for fuel failure was deduced using Monte Carlo method. The detection sensitivity can be defined by the ratio of the amount of U-235 exposed by the coolant and the neutron count rate, i.e., the U-235 exposure per unit delayed neutron count rate. If we determine this sensitivity, we can estimate the seriousness of the fuel failure, and set the criteria to judge the reactor operation with measured neutron count rate.
1135
8000
Discrimination level 6000 4000 2000 0 0
90
180
270
360
450
Channel number Fig. 1. Pulse height spectrum from the BF3 detector installed around the primary cooling pipe of HANARO.
order to decrease the N-16 effect. In this experiment, the time required for the outgoing coolant from the reactor core to reach the detector position was 4.7 s. This coolant passed by the detector again 15.6 s later through the inlet pipe. The circulation time of the coolant was 21.4 s. During normal operation of HANARO, the possible neutron sources around the primary cooling pipe are the delayed neutrons from the uranium contaminated on the fuel surface, and the photoneutrons from the deuterium in the coolant. The delayed neutrons can be generated from the failed fuel as well as surface contamination in the abnormal situation for HANARO fuel. Since the uranium inside the fuel undergoes fission in the same way as contaminated one, these two sources cannot be separated. However, in this experiment, the established fuel failure detection system of HANARO and the fuel inspection process did not detect any defect in fuels. Therefore, we assumed that the experiments were performed in normal operation of reactor, and that all delayed neutrons were from surface contamination. The delayed neutron generation due to the surface contamination does not decrease in time of reactor operation in spite of the forced coolant circulation because it comes from fixed contamination that has not been eliminated in the thorough washing out process after fabrication of the fuel. From the quantitative analysis of the radionuclide in the primary coolant of HANARO using HPGe detector, it was confirmed that the Na-24 activity concentration in the coolant was relatively small. Therefore, the high energy gamma-rays emitted from the N-16 are the major source for the photoneutron production. If the photoneutron contribution to the total neutron count is large, the neutron detection system may not be so sensitive for the detection of the fuel failure. In order to find out the portion of the photoneutrons and the delayed neutrons, the difference of the half-lives between N-16 and delayed neutron precursors was utilized.
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
N 0 ðtÞ ¼ GðtÞ þ N i ðtÞ:
ð1Þ
Since the residence time of the coolant in the core region is very short compared with the half-life of N-16, the decay of N-16 in the core can be neglected. If the coolant transit time from the core outlet to the inlet through the primary cooling circuit is Ts, then Ni is N i ðtÞ ¼
M N 0 ðt T s Þ exp ðkT s Þ; M þm
0
10
Measured Calculated [Delayed neutron] 10-1
10-2
Reactor power
ð3Þ
The delayed neutron generation rate in the primary coolant circuit can also be obtained by considering the generations and decays of the delayed neutron precursors with different half-lives as above. The variation of the neutron count rate during a sudden trip of the reactor was measured together with the power variation. BF3 detectors were installed below the coolant outlet and neighboring inlet pipe. The power variation was measured using compensated ion chamber (CIC). The relative variations of the photoneutron and delayed neutron count rates according to the power history after reactor trip were calculated, and they were compared with the neutron measurements. Fig. 2 shows the measured and calculated results for the changes of the photoneutrons and the delayed neutrons before and after reactor trip. The steps in calculation curves were represented because we assumed that the coolant was not mixed and that it had a constant velocity. In Fig. 2a, the measured count rate of BF3 detector is decreased rapidly after about 4 s from the reactor shutdown because of the delayed arrival of the coolant. The first drop of the measurements is similar with the calculated values. However, from the second drop, the number of photoneutrons decreases much faster than that of the delayed neutrons. The measured values are closer to the delayed neutron calculation but they are between the calculated values of the delayed neutron and photoneutron. Fig. 2b shows the changes of the neutron count rates for the inlet pipe. The measurements show the first drop at the same time as the core outlet case, which is not found in the calculations. The calculations show much larger first drops than the measurements at about 15 s after reactor trip. It indicates that the neutron measurements at the core inlet are significantly influenced from the neutrons generated
Calculated [Photoneutron]
10-3 0
ð2Þ
where k is the decay constant of N-16, M and m are the coolant flow rates in the primary cooling circuit and in the bypass loop for purification, respectively. If the coolant transit time from the core outlet to the BF3 detector position is Td, the specific activity at this position becomes N ðtÞ ¼ N 0 ðt T d Þ exp ðkT d Þ:
a
Relative unit
If the generation rate of N-16 by O16(n, p)N16 reaction in the unit coolant volume of reactor core is G, and the specific activity of N-16 in the coolant coming into the core is Ni, then the specific activity of N-16 in the coolant leaving the core at time t, N0 is given by
50
100
150
200
Time [sec]
b
0
10
Measured Calculated [Delayed neutron]
Relative unit
1136
10
-1
Calculated [Photoneutron]
Reactor power
10-2
10-3 0
50
100
150
200
Time [sec] Fig. 2. The changes of the measured neutron count rates and calculated values for photoneutron and delayed neutron below the core outlet (a) and inlet (b) pipes before and after reactor trip.
at the nearby core outlet pipes. After the first drop, the measurements are between two calculations. It is evident that the measured neutrons consist of photoneutrons and delayed neutrons coming from both of core inlet and outlet pipes. Each contribution was investigated by fitting the calculated values to the measurements, and the results are depicted in Fig. 3. For the case of the detector at the core outlet pipe, it was assumed that 98% of the neutron counts were from the core outlet pipe, and that 2% were from the core inlet pipe. For the detector at the inlet pipe, the contributions from the inlet and outlet pipes were assumed to be 50% each. And also, it was assumed that 30% of the neutrons generated from the outlet pipe were photoneutrons, and that the remainder were delayed neutrons. In order to enhance the sensitivity of neutron detection, the polyethylene reflector was arranged around the BF3 detector (Jacobi et al., 1977). In that case, the neutron count rate was greatly enhanced by about three times, i.e. 900 cps, and this count rate was high enough for stable signal to monitor the fuel failure with short counting time.
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
from O16(n, p)N16 reaction (T1/2 = 7.1 s), and their energies are 6.13 MeV(69%) and 7.12 MeV(5%), respectively. We assumed that these gamma-rays could contribute to the generation of photoneutron. The cross-section of O16(n, p)N16 reaction averaged for the fission spectrum is 3.49 · 105 barns (NEA-OECD, 1994). The saturated specific activity of N-16 at the reactor core outlet can be written by
0
Relative unit
10
Reactor power
-1
10
Core inlet
-2
10
Core outlet
: Measured : Calculated
-3
10
0
50
1 ekti 1 ekðti þt0 Þ Z 1 ekti ¼ N O rðEÞ/ðEÞ dE; 1 ekðti þt0 Þ
AN ¼ P N 100
150
200
Time [sec] Fig. 3. The measured neutron count rates and the calculated results obtained by settling the effects of core inlet and outlet pipes and the contributions of the photoneutrons and the delayed neutrons to the detectors.
3. Calculation of neutron count In order to not only confirm the reliability of the measurements but also estimate the sensitivity of the fuel failure detection, the neutron count rates by BF3 detector at core outlet pipe were predicted by the Monte Carlo calculation. We made the peculiar Monte Carlo program applicable to this problem, and the generation, transportation and detection of the photoneutrons and the delayed neutrons were simulated for the geometry similar to the experiments. The calculation model with polyethylene reflector is shown in Fig. 4. The thicknesses of two lead layers and a polyethylene layer are 10 cm each. It is assumed that the coolant pipe is surrounded by concrete with infinite thickness, which is 1 m apart from center of pipe. The neutron detector is put into the polyethylene groove. The photoneutron and delayed neutron fluxes at this detector location were obtained, respectively. The length and diameter of primary cooling pipe viewed by the detector were 700 and 33.02 cm. The thickness of pipe was neglected. Using the calculated neutron flux and the energy dependent efficiency of the BF3 detector used in the experiments, the neutron count rates were obtained. The main gamma-rays at the HANARO primary cooling circuit originate from the decays of N-16 generated
Lead
Coolant pipe
1137
Detector
ð4Þ
where PN is the N-16 production rate, NO is the number density of O-16 nucleus in coolant, /(E) is the energy dependent neutron flux, r(E) is the energy dependent microscopic reaction cross-section, ti and t0 are the passing times of the coolant through the core and primary cooling circuit, respectively, and k is the decay constant of N-16. The coolant in the HANARO core can flow through the flow tube where the nuclear fuels are located and other paths like the gap between the flow tubes. The N-16 production rate in the coolant of the core was obtained from the calculation of the O16(n, p)N16 reaction rate using MCNP for the same core condition as experiment. For the reactor power of 20 MW, the specific activity of N-16 at the core outlet was calculated to be 2.03 · 106 dps. A gamma-ray bunch was generated at a position within the coolant pipe viewed by the BF3 detector. This bunch corresponds to the coolant with unit volume, and the number of gamma-rays in a bunch was obtained from the specific activity of N-16 at the core outlet and its decay. The generation position of a bunch in the coolant pipe and the energy and the direction of the gamma-rays from the bunch were determined by random numbers, which were obtained by the recurrence equation (Press et al., 1994). The period of the random number was 231–2 = 2.1 · 109. The number of gamma-ray bunch histories was set so that the total volume of the generated bunch might be the same as the total coolant volume viewed by the detector. For each history of gamma-ray bunch, the types and the positions of interactions, and the directions and energies of secondary particles were generated, and then, the amount and position of generated photoneutron were simulated. The moving distance t to the next interaction position of the gamma-ray in a medium with a total macroscopic cross-section Rt can be extracted from R t R t0 e t Rt dt0 0 R ¼ R1 ; ð5Þ eRt t0 Rt dt0 0 as follows
Concrete Polyethylene Fig. 4. The model used in the transport calculations of photoneutrons and delayed neutrons.
t¼
1 ln R; Rt
where R is a random number.
ð6Þ
1138
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
In the gamma-ray interactions, the photoelectric effect was neglected because its cross-section was less than 0.05% of the cross-sections of Compton scattering and pair production in the several MeV energy range. The cross-sections of the main interaction processes of the gamma-rays were taken from the Photon Cross Section Database (Berger et al., 2005). The polar angle after Compton scattering was calculated from the Klein–Nishina formula. The azimuthal angle was set by uniform sampling between 0 and 2p because the photon polarization was neglected (Bak et al., 1995). In the transport of a gamma-ray bunch, when the pair production had occurred, or gamma-ray energy had gone below 2.225 MeV that is the binding energy of the deuteron, the calculation loop was stopped. The photodisintegration interaction of deuteron can occur at a random position of moving path of the gamma-ray in the coolant water. The number of generated photoneutrons was determined from the number of gamma-rays in the bunch which was being transported, the path length of the bunch in the coolant water, the number density of the deuteron in the water and the photodisintegration cross-section. The photodisintegration cross-section was deduced from the cross-sections of magnetic and electric dipole disintegrations of the deuteron (Blatt and Weisskopf, 1979). The angular distribution of generated neutron was assumed to be isotropic. In the neutron transport in matter, the motion of target nucleus, chemical bonding, crystal lattice effects and inelastic scattering were neglected, and the elastic scattering occurred isotropically in center of mass system. The neutron crosssection used in this calculation was taken from ENDF/ B6.1 (Rose et al., 1991). The neutron energy after isotropic elastic scattering can be written by E ¼ E0 ½a þ ð1 aÞR;
ð7Þ
where E0 is the initial neutron energy, R is a random number, and, 2 A1 ; ð8Þ a¼ Aþ1 where A is the nucleon number of target nucleus. The scattering angle in laboratory system was obtained from rffiffiffiffiffi rffiffiffiffiffi Aþ1 E A1 E0 cos h ¼ : ð9Þ 2 E0 2 E The azimuthal angle of the scattered neutron was determined in the same way as that of Compton scattered photon. The neutron going out of the coolant pipe experienced the new media such as lead, polyethylene and concrete. The neutron coming into detector region experienced the neutron detection process by BF3 detector dependent on its energy and path length (Gardner et al., 1996; Knoll, 2000).
The simulation process for the transport of the delayed neutron is very similar to that of the photoneutron case except the generation procedure. The rod-type metallic fuels are used in HANARO. The fuel meat consists of the dispersed small particles of a high density uranium silicide compound in a continuous aluminum matrix. The reference fuel material has a nominal composition of 61.4 w/o U3Si and 38.6 w/o Al. The cladding is the reactor grade aluminum. To increase the heat transfer area and to provide the stiffness for the elements, the fuel elements use the finned cladding. The inside surface of the cladding is in intimate contact with the fuel meat as a result of the extrusion cladding process. The grain size of the uranium silicide compound inside the fuel meat is several tens of lm. Because there is no gap between the fuel meat and the cladding, when some defect occurs in the fuel rod, the fuel meat meets the main coolant flow. Thus, the fission product can leave the fuel to enter the coolant. The release of the neutron emitting fission products depends on the total number of the produced nuclides, effective leakage area and the recoil length (Lustman, 1961; Jacobi et al., 1977). In the normal operation of the reactor, the delayed neutron is generated from the contamination of the uranium silicide grains on the fuel surface. In order to determine the relation between the fission rate in the flow tube by the surface contamination and the detected count rate of the delayed neutron, we assumed that 1.0 · 105 fissions of U-235 nuclei occurred in the unit coolant volume of the reactor core. From the microscopic analysis, it was confirmed that the pathlength of the fission product required to escape from a grain is mostly smaller than the range of the fission product (Bridwell and Moak, 1967). Therefore, it was also assumed that all of the generated delayed neutron precursors were transferred by the coolant. The delayed neutrons generated from the decay of the precursor were transported in the same manner as the transportation of the photoneutrons. The above transport procedures for the photoneutrons and delayed neutrons were performed with or without polyethylene reflector. The measured and calculated neutron count rates around the primary coolant pipe are represented in Table 1. From the table, it is confirmed that the ratios of the calculated count rates for photoneutron and delayed neutron with polyethylene to those without it are similar to that of measurements. This may support the reliability and sensitivity of the calculation. The calculated count rate of photoneutron is deduced to be about 20% of measured neutron count rate. In the experiments, it was predicted to be about 30%. This difference could originate from several sources such as uncertainty in the deduction process of experimental result, geometrical difference between the calculation and the experiment, uncertainties in the detection efficiency and the MCNP calculation, activation of water in no fuel region. However, the difference is rather small,
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
1139
Table 1 Measured and calculated neutron count rates around the primary coolant pipe
Measured count rate (cps) Calculated photoneutron count rate (cps) Delayed neutron count rate calculated by assuming 1.0 · 105 fission in unit coolant volume (cps)
With polyethylene reflector (A)
Without polyethylene reflector (B)
B/A
900 180
280 50
0.31 0.28
620
190
0.31
considering the uncertainties of the above parameters and methods in the calculation and experiment. 4. Sensitivity of fuel failure detection If we regard the calculated photoneutron count rate in Table 1 as the real value at the experiment, we can deduce the delayed neutron count rate in the measurements, and it becomes 720 cps. Since the delayed neutron count rate of 620 cps was obtained by assuming 1.0 · 105 fissions of U235 nuclei in the unit coolant volume of the core, the real number of fission at the experiment can be estimated to be 1.16 · 105 fissions/cm3 s as shown in Table 2. While the fissions occurred in the flow tube, where the nuclear fuels were located, the coolant water going out of the reactor core contained the water through the flow tube and other paths. Since the ratio of the amount of flowing water through the flow tube to that of the core passing coolant is 0.7, the fission rate in the flow tube by surface contamination is calculated to be 1.66 · 105 fissions/cm3 s. In order to estimate the sensitivity of U-235 detection by the delayed neutron measurements, the amount of uranium exposed to the coolant in the core was also predicted. The fission rate of 1 lg U-235 uniformly contaminated on the surface of the fuel rods and exposed to the coolant at the flow tube was calculated by MCNP. The core condition in this calculation was the same as that in the calculation of N-16 generation rate. The calculated value of U-235 fission rate in the flow tube is 8.99 · 107 fissions/lg-U235 s. Using this value Table 2 Deduced delayed neutron count rates and fission rate in the flow tube by surface contamination Delayed neutron count rate (cps) Deduced from measurements and photoneutron calculation (a) Calculated by assuming 1.0 · 105 fissions (b) Fission rate of U-235 in unit coolant volume of the core (1.0 · 105 · a/b) (fissions/cm3 s) Volume ratio of coolant through flow tube to total core outgoing water Deduced fission rate in the flow tube by surface contamination (fissions/cm3 s)
720 620 1.16 · 105
0.7
1.66 · 105
and the estimated fission rate in the flow tube, we can obtain the mass of U-235 which was exposed to the coolant in the flow tube, and it is 1.85 · 103lg/cm3. Considering the total volume of flow tube, the mass of U-235 that was exposed to the coolant in the experiment can be estimated to be 120 lg. The maximum allowable surface contamination of HANARO fuel is 3.25 lg-U235/ft2, i.e., the total allowable mass of contaminated U-235 on the surface of fuel rods loaded in HANARO is 940 lg. Therefore, it is confirmed that the count rate of delayed neutron measured in the primary cooling circuit may have originated from about 13% of the maximum allowable surface contamination. From above results, the sensitivity of U-235 mass detection by delayed neutron measurements can be estimated to be about 0.2 lg-U235/cps. The delayed neutron count rate of about 720 cps at 20 MW reactor power can be obtained from the U-235 exposure of 120 lg, which is much lower than the allowable level of surface contamination of fuel rod. And, the contributions of other components such as photoneutron and gamma-rays are rather small. Therefore, the delayed neutron detection method is sensitive enough to detect the fuel failure, and the neutron count rate is high enough for stable signal with short counting time. In abnormal condition with the fuel failure, the delayed neutron count rates from the failed fuel will be added to the background count rates due to the delayed neutron from the surface contamination and the photoneutron. By using above sensitivity and increased neutron count rate, we can estimate the seriousness of the fuel failure and set the criteria for the safe reactor operation. 5. Conclusion The sensitivity of the fuel failure detection based on the delayed neutron measurement in the primary cooling circuit of HANARO reactor is investigated by both calculation and measurement. The neutron count rate of 900 cps at 20 MW reactor power can be obtained, and about 720 cps is estimated to be the contribution of delayed neutrons. It is confirmed that the delayed neutron count rate can be obtained from the U-235 exposure of 120 lg, which is much lower than the allowable level of surface contamination of fuel rod. The contributions of other components such as photoneutron and gamma-rays are
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rather small. Conclusively, the delayed neutron detection method is sensitive enough to detect fuel failure, and that the neutron count rate is high enough for stable signal with short counting time. References Bak, H.I., Bae, Y.D., Kim, M.S., Choi, H.D., 1995. Peak energy shift dependent on source position in c-ray energy measurement by using a closed-ended coaxial HPGe detector. Nucl. Instrum. Meth. A366, 332– 339. Berger, M.J., Hubbell, J.H., Seltzer, S.M., Chang, J., Coursey, J.S., Sukumar, R., Zucker, D.S., 2005. Downloading Data from INTERNET. Available from: http://www.physics.nist.gov/PhysRefData/ Xcom/Text/XCOM.html. Blatt, J.M., Weisskopf, V.F., 1979. Theoretical Nuclear Physics. SpringerVerlag. Bridwell, L.B., Moak, C.D., 1967. Stopping powers and differential ranges for 79Br and 127I in UF4. Phys. Rev. 156, 242–243. Dobrin, R., Craciunescu, T., Tuturici, I.L., 1997. The analysis of failed nuclear fuel rods by gamma computed tomography. J. Nucl. Mat. 246, 37–42. Gardner, R.P., Barrett, C.L., Haq, W., Peplow, D.E., 1996. Efficient Monte Carlo simulation of 16O neutron activation and 16N decay
gamma-ray detection in a flowing fluid for on-line oxygen analysis or flow rate measurement. Nucl. Sci. Eng. 122, 326–343. Jacobi, S., Letz, K., Schmitz, G., 1977. Release and detection of fission products from defective fuel pins. Nucl. Eng. Des. 44, 125– 134. Kawai, M., Hayashida, Y., Nishi, H., 1990. Application of forward and adjoint Monte Carlo coupling technique to detector system designs. Prog. Nucl. Energ. 24, 311–319. Knoll, G.F., 2000. Radiation Detection and Measurement. John Wiley & Sons, New York. Lustman, B., 1961. Irradiation effects in uranium dioxide. In: Belle, J. (Ed.), Uranium Dioxide: Properties and Nuclear Applications. US Government Printing Office, Washington, DC, p. 481. NEA-OECD, 1994. Table of Simple Integral Neutron Cross Section Data from JEF-2.2, ENDF/B-VI, JENDL-3.2, BROND-2 and CENDL-2, JEF Report 14, Paris. Press, W.H., Teukolsky, S.A., Vetterling, W.T., Flannery, B.P., 1994. Numerical Recipes. Cambridge University Press, Cambridge. Rose, P.F. (Ed.), Cross Section Evaluation Working Group, National Nuclear Data Center, Brookhaven National Laboratory, 1991. ENDF/B-VI Summary Documentation, Report BNL-NCS-17541 (ENDF-201). Venkateswaran, G., Venkateswarlu, K.S., 1989. Theoretical assessment of ´ -VIS channel DN monitoring for channel fission gas monitoring VIS-A PHWRs. Ann. Nucl. Energ. 16, 139–149.
NUCLEAR ENERGY Annals of Nuclear Energy 33 (2006) 1134–1140 www.elsevier.com/locate/anucene
Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO Myong-Seop Kim *, Byung-Chul Lee, Sang-Jun Park, Byung-Jin Jun Korea Atomic Energy Research Institute, 150, Deokjin-Dong, Yuseong, Daejeon 305-353, Republic of Korea Received 10 March 2006; received in revised form 23 June 2006; accepted 23 June 2006 Available online 28 August 2006
Abstract The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF3 detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF3 detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 · 105 fissions/ cm3 s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 lg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 lg-U235/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time. 2006 Elsevier Ltd. All rights reserved.
1. Introduction In research reactors as well as nuclear power plants, any fuel failure that causes a release of fission products into the primary coolant should be detected as soon as possible. In HANARO, 30 MW research reactor, the fuel failure had been monitored using the gamma-ray detection system. Three NaI(Tl) detectors were placed under the common return line of the primary cooling circuit in order to monitor the gamma-ray activity in the coolant caused by the release of the fission products. From the operation experiences and various analyses, it was ascertained that the fuel failure detection system
*
Corresponding author. Tel.: +82 42 868 8666; fax: +82 42 868 8341. E-mail address: [email protected] (M.S. Kim).
0306-4549/$ - see front matter 2006 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2006.06.002
(FFDS) using NaI(Tl) detector was facing difficulties in monitoring the fuel failure because of N-16, a major gamma-ray source in the coolant. The energies of the gamma-rays from the fission and activation products except N-16 are mostly smaller than 2 MeV. Measuring the total gamma-ray activity below 2 MeV was not sensitive enough to detect the fuel failure due to the high Compton background to the detector of the high energy gamma-rays from the N-16. If a decay system is installed to decrease the N-16 effect, the detection of fuel failure would be delayed in time. Therefore, in HANARO, the N-16 effect had been compensated for by using the predetermined Compton background and the ratio of the net count rate of the gamma-rays with 0.6–2.0 MeV to that above 2.0 MeV. This method provided the sensitive detection of the fuel failure, but the net count rate was not so stable because of the statistical fluctuation. Also, the ratio
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
2. Neutron measurements and verification of their origins In order to measure the neutrons around the primary cooling circuit of HANARO, BF3 proportional detectors were used. The effect of the high energy gamma-rays on the neutron pulse spectrum of the BF3 detector was investigated. The diameter of the BF3 detector is 25 mm, and its sensitive axial length is 120 mm. Fig. 1 shows the pulse height spectrum from the BF3 detector installed below the coolant outlet pipe. Since the neutron pulses are well discriminated from the gamma-ray pulses and the electronic noises, the effect of high energy gamma-rays on the neutron pulse spectrum of the BF3 detector is negligible. Therefore, the BF3 detector could be installed at the position as close as possible to the coolant outlet from the core, i.e., the delaying in time for detection of fuel failure was minimized. The time sensitivity of this system was improved much more than that of previous FFDS using NaI(Tl) detector that should be installed around the pipe of the return line in
12000 10000
Counts/channel
was rapidly increased when the reactor power was being decreased because the decay of the activation products was slower than that of the N-16, which caused an unscheduled reactor trip. Another possibility to monitor the fuel failure is the detection of the delayed neutrons from the short-lived fission products such as Br-87 and I-137 (Dobrin et al., 1997; Jacobi et al., 1977). These delayed neutron precursors are released to the coolant by a recoil process from the inside of the failed fuel rods (Venkateswaran and Venkateswarlu, 1989). During normal reactor operation, the neutrons can be generated in and out of primary cooling circuit from various sources such as streaming through the coolant duct, photodisintegration by high energy gamma-rays from activated nuclei like N-16 and Na-24, and delayed neutron generation from contaminated uranium on fuel surface (Jacobi et al., 1977; Kawai et al., 1990; Venkateswaran and Venkateswarlu, 1989). In order to apply the FFDS based on the delayed neutron measurement to HANARO operation, it should be confirmed that the rapid and sensitive detection of neutron from the inside of the failed fuel rod is possible regardless of the background contributions of above neutrons. Therefore, in this work, the neutrons from the primary coolant during the normal operation of HANARO were measured, and the origins of these neutrons were analyzed. Then, the detection sensitivity for fuel failure was deduced using Monte Carlo method. The detection sensitivity can be defined by the ratio of the amount of U-235 exposed by the coolant and the neutron count rate, i.e., the U-235 exposure per unit delayed neutron count rate. If we determine this sensitivity, we can estimate the seriousness of the fuel failure, and set the criteria to judge the reactor operation with measured neutron count rate.
1135
8000
Discrimination level 6000 4000 2000 0 0
90
180
270
360
450
Channel number Fig. 1. Pulse height spectrum from the BF3 detector installed around the primary cooling pipe of HANARO.
order to decrease the N-16 effect. In this experiment, the time required for the outgoing coolant from the reactor core to reach the detector position was 4.7 s. This coolant passed by the detector again 15.6 s later through the inlet pipe. The circulation time of the coolant was 21.4 s. During normal operation of HANARO, the possible neutron sources around the primary cooling pipe are the delayed neutrons from the uranium contaminated on the fuel surface, and the photoneutrons from the deuterium in the coolant. The delayed neutrons can be generated from the failed fuel as well as surface contamination in the abnormal situation for HANARO fuel. Since the uranium inside the fuel undergoes fission in the same way as contaminated one, these two sources cannot be separated. However, in this experiment, the established fuel failure detection system of HANARO and the fuel inspection process did not detect any defect in fuels. Therefore, we assumed that the experiments were performed in normal operation of reactor, and that all delayed neutrons were from surface contamination. The delayed neutron generation due to the surface contamination does not decrease in time of reactor operation in spite of the forced coolant circulation because it comes from fixed contamination that has not been eliminated in the thorough washing out process after fabrication of the fuel. From the quantitative analysis of the radionuclide in the primary coolant of HANARO using HPGe detector, it was confirmed that the Na-24 activity concentration in the coolant was relatively small. Therefore, the high energy gamma-rays emitted from the N-16 are the major source for the photoneutron production. If the photoneutron contribution to the total neutron count is large, the neutron detection system may not be so sensitive for the detection of the fuel failure. In order to find out the portion of the photoneutrons and the delayed neutrons, the difference of the half-lives between N-16 and delayed neutron precursors was utilized.
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
N 0 ðtÞ ¼ GðtÞ þ N i ðtÞ:
ð1Þ
Since the residence time of the coolant in the core region is very short compared with the half-life of N-16, the decay of N-16 in the core can be neglected. If the coolant transit time from the core outlet to the inlet through the primary cooling circuit is Ts, then Ni is N i ðtÞ ¼
M N 0 ðt T s Þ exp ðkT s Þ; M þm
0
10
Measured Calculated [Delayed neutron] 10-1
10-2
Reactor power
ð3Þ
The delayed neutron generation rate in the primary coolant circuit can also be obtained by considering the generations and decays of the delayed neutron precursors with different half-lives as above. The variation of the neutron count rate during a sudden trip of the reactor was measured together with the power variation. BF3 detectors were installed below the coolant outlet and neighboring inlet pipe. The power variation was measured using compensated ion chamber (CIC). The relative variations of the photoneutron and delayed neutron count rates according to the power history after reactor trip were calculated, and they were compared with the neutron measurements. Fig. 2 shows the measured and calculated results for the changes of the photoneutrons and the delayed neutrons before and after reactor trip. The steps in calculation curves were represented because we assumed that the coolant was not mixed and that it had a constant velocity. In Fig. 2a, the measured count rate of BF3 detector is decreased rapidly after about 4 s from the reactor shutdown because of the delayed arrival of the coolant. The first drop of the measurements is similar with the calculated values. However, from the second drop, the number of photoneutrons decreases much faster than that of the delayed neutrons. The measured values are closer to the delayed neutron calculation but they are between the calculated values of the delayed neutron and photoneutron. Fig. 2b shows the changes of the neutron count rates for the inlet pipe. The measurements show the first drop at the same time as the core outlet case, which is not found in the calculations. The calculations show much larger first drops than the measurements at about 15 s after reactor trip. It indicates that the neutron measurements at the core inlet are significantly influenced from the neutrons generated
Calculated [Photoneutron]
10-3 0
ð2Þ
where k is the decay constant of N-16, M and m are the coolant flow rates in the primary cooling circuit and in the bypass loop for purification, respectively. If the coolant transit time from the core outlet to the BF3 detector position is Td, the specific activity at this position becomes N ðtÞ ¼ N 0 ðt T d Þ exp ðkT d Þ:
a
Relative unit
If the generation rate of N-16 by O16(n, p)N16 reaction in the unit coolant volume of reactor core is G, and the specific activity of N-16 in the coolant coming into the core is Ni, then the specific activity of N-16 in the coolant leaving the core at time t, N0 is given by
50
100
150
200
Time [sec]
b
0
10
Measured Calculated [Delayed neutron]
Relative unit
1136
10
-1
Calculated [Photoneutron]
Reactor power
10-2
10-3 0
50
100
150
200
Time [sec] Fig. 2. The changes of the measured neutron count rates and calculated values for photoneutron and delayed neutron below the core outlet (a) and inlet (b) pipes before and after reactor trip.
at the nearby core outlet pipes. After the first drop, the measurements are between two calculations. It is evident that the measured neutrons consist of photoneutrons and delayed neutrons coming from both of core inlet and outlet pipes. Each contribution was investigated by fitting the calculated values to the measurements, and the results are depicted in Fig. 3. For the case of the detector at the core outlet pipe, it was assumed that 98% of the neutron counts were from the core outlet pipe, and that 2% were from the core inlet pipe. For the detector at the inlet pipe, the contributions from the inlet and outlet pipes were assumed to be 50% each. And also, it was assumed that 30% of the neutrons generated from the outlet pipe were photoneutrons, and that the remainder were delayed neutrons. In order to enhance the sensitivity of neutron detection, the polyethylene reflector was arranged around the BF3 detector (Jacobi et al., 1977). In that case, the neutron count rate was greatly enhanced by about three times, i.e. 900 cps, and this count rate was high enough for stable signal to monitor the fuel failure with short counting time.
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
from O16(n, p)N16 reaction (T1/2 = 7.1 s), and their energies are 6.13 MeV(69%) and 7.12 MeV(5%), respectively. We assumed that these gamma-rays could contribute to the generation of photoneutron. The cross-section of O16(n, p)N16 reaction averaged for the fission spectrum is 3.49 · 105 barns (NEA-OECD, 1994). The saturated specific activity of N-16 at the reactor core outlet can be written by
0
Relative unit
10
Reactor power
-1
10
Core inlet
-2
10
Core outlet
: Measured : Calculated
-3
10
0
50
1 ekti 1 ekðti þt0 Þ Z 1 ekti ¼ N O rðEÞ/ðEÞ dE; 1 ekðti þt0 Þ
AN ¼ P N 100
150
200
Time [sec] Fig. 3. The measured neutron count rates and the calculated results obtained by settling the effects of core inlet and outlet pipes and the contributions of the photoneutrons and the delayed neutrons to the detectors.
3. Calculation of neutron count In order to not only confirm the reliability of the measurements but also estimate the sensitivity of the fuel failure detection, the neutron count rates by BF3 detector at core outlet pipe were predicted by the Monte Carlo calculation. We made the peculiar Monte Carlo program applicable to this problem, and the generation, transportation and detection of the photoneutrons and the delayed neutrons were simulated for the geometry similar to the experiments. The calculation model with polyethylene reflector is shown in Fig. 4. The thicknesses of two lead layers and a polyethylene layer are 10 cm each. It is assumed that the coolant pipe is surrounded by concrete with infinite thickness, which is 1 m apart from center of pipe. The neutron detector is put into the polyethylene groove. The photoneutron and delayed neutron fluxes at this detector location were obtained, respectively. The length and diameter of primary cooling pipe viewed by the detector were 700 and 33.02 cm. The thickness of pipe was neglected. Using the calculated neutron flux and the energy dependent efficiency of the BF3 detector used in the experiments, the neutron count rates were obtained. The main gamma-rays at the HANARO primary cooling circuit originate from the decays of N-16 generated
Lead
Coolant pipe
1137
Detector
ð4Þ
where PN is the N-16 production rate, NO is the number density of O-16 nucleus in coolant, /(E) is the energy dependent neutron flux, r(E) is the energy dependent microscopic reaction cross-section, ti and t0 are the passing times of the coolant through the core and primary cooling circuit, respectively, and k is the decay constant of N-16. The coolant in the HANARO core can flow through the flow tube where the nuclear fuels are located and other paths like the gap between the flow tubes. The N-16 production rate in the coolant of the core was obtained from the calculation of the O16(n, p)N16 reaction rate using MCNP for the same core condition as experiment. For the reactor power of 20 MW, the specific activity of N-16 at the core outlet was calculated to be 2.03 · 106 dps. A gamma-ray bunch was generated at a position within the coolant pipe viewed by the BF3 detector. This bunch corresponds to the coolant with unit volume, and the number of gamma-rays in a bunch was obtained from the specific activity of N-16 at the core outlet and its decay. The generation position of a bunch in the coolant pipe and the energy and the direction of the gamma-rays from the bunch were determined by random numbers, which were obtained by the recurrence equation (Press et al., 1994). The period of the random number was 231–2 = 2.1 · 109. The number of gamma-ray bunch histories was set so that the total volume of the generated bunch might be the same as the total coolant volume viewed by the detector. For each history of gamma-ray bunch, the types and the positions of interactions, and the directions and energies of secondary particles were generated, and then, the amount and position of generated photoneutron were simulated. The moving distance t to the next interaction position of the gamma-ray in a medium with a total macroscopic cross-section Rt can be extracted from R t R t0 e t Rt dt0 0 R ¼ R1 ; ð5Þ eRt t0 Rt dt0 0 as follows
Concrete Polyethylene Fig. 4. The model used in the transport calculations of photoneutrons and delayed neutrons.
t¼
1 ln R; Rt
where R is a random number.
ð6Þ
1138
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In the gamma-ray interactions, the photoelectric effect was neglected because its cross-section was less than 0.05% of the cross-sections of Compton scattering and pair production in the several MeV energy range. The cross-sections of the main interaction processes of the gamma-rays were taken from the Photon Cross Section Database (Berger et al., 2005). The polar angle after Compton scattering was calculated from the Klein–Nishina formula. The azimuthal angle was set by uniform sampling between 0 and 2p because the photon polarization was neglected (Bak et al., 1995). In the transport of a gamma-ray bunch, when the pair production had occurred, or gamma-ray energy had gone below 2.225 MeV that is the binding energy of the deuteron, the calculation loop was stopped. The photodisintegration interaction of deuteron can occur at a random position of moving path of the gamma-ray in the coolant water. The number of generated photoneutrons was determined from the number of gamma-rays in the bunch which was being transported, the path length of the bunch in the coolant water, the number density of the deuteron in the water and the photodisintegration cross-section. The photodisintegration cross-section was deduced from the cross-sections of magnetic and electric dipole disintegrations of the deuteron (Blatt and Weisskopf, 1979). The angular distribution of generated neutron was assumed to be isotropic. In the neutron transport in matter, the motion of target nucleus, chemical bonding, crystal lattice effects and inelastic scattering were neglected, and the elastic scattering occurred isotropically in center of mass system. The neutron crosssection used in this calculation was taken from ENDF/ B6.1 (Rose et al., 1991). The neutron energy after isotropic elastic scattering can be written by E ¼ E0 ½a þ ð1 aÞR;
ð7Þ
where E0 is the initial neutron energy, R is a random number, and, 2 A1 ; ð8Þ a¼ Aþ1 where A is the nucleon number of target nucleus. The scattering angle in laboratory system was obtained from rffiffiffiffiffi rffiffiffiffiffi Aþ1 E A1 E0 cos h ¼ : ð9Þ 2 E0 2 E The azimuthal angle of the scattered neutron was determined in the same way as that of Compton scattered photon. The neutron going out of the coolant pipe experienced the new media such as lead, polyethylene and concrete. The neutron coming into detector region experienced the neutron detection process by BF3 detector dependent on its energy and path length (Gardner et al., 1996; Knoll, 2000).
The simulation process for the transport of the delayed neutron is very similar to that of the photoneutron case except the generation procedure. The rod-type metallic fuels are used in HANARO. The fuel meat consists of the dispersed small particles of a high density uranium silicide compound in a continuous aluminum matrix. The reference fuel material has a nominal composition of 61.4 w/o U3Si and 38.6 w/o Al. The cladding is the reactor grade aluminum. To increase the heat transfer area and to provide the stiffness for the elements, the fuel elements use the finned cladding. The inside surface of the cladding is in intimate contact with the fuel meat as a result of the extrusion cladding process. The grain size of the uranium silicide compound inside the fuel meat is several tens of lm. Because there is no gap between the fuel meat and the cladding, when some defect occurs in the fuel rod, the fuel meat meets the main coolant flow. Thus, the fission product can leave the fuel to enter the coolant. The release of the neutron emitting fission products depends on the total number of the produced nuclides, effective leakage area and the recoil length (Lustman, 1961; Jacobi et al., 1977). In the normal operation of the reactor, the delayed neutron is generated from the contamination of the uranium silicide grains on the fuel surface. In order to determine the relation between the fission rate in the flow tube by the surface contamination and the detected count rate of the delayed neutron, we assumed that 1.0 · 105 fissions of U-235 nuclei occurred in the unit coolant volume of the reactor core. From the microscopic analysis, it was confirmed that the pathlength of the fission product required to escape from a grain is mostly smaller than the range of the fission product (Bridwell and Moak, 1967). Therefore, it was also assumed that all of the generated delayed neutron precursors were transferred by the coolant. The delayed neutrons generated from the decay of the precursor were transported in the same manner as the transportation of the photoneutrons. The above transport procedures for the photoneutrons and delayed neutrons were performed with or without polyethylene reflector. The measured and calculated neutron count rates around the primary coolant pipe are represented in Table 1. From the table, it is confirmed that the ratios of the calculated count rates for photoneutron and delayed neutron with polyethylene to those without it are similar to that of measurements. This may support the reliability and sensitivity of the calculation. The calculated count rate of photoneutron is deduced to be about 20% of measured neutron count rate. In the experiments, it was predicted to be about 30%. This difference could originate from several sources such as uncertainty in the deduction process of experimental result, geometrical difference between the calculation and the experiment, uncertainties in the detection efficiency and the MCNP calculation, activation of water in no fuel region. However, the difference is rather small,
M.S. Kim et al. / Annals of Nuclear Energy 33 (2006) 1134–1140
1139
Table 1 Measured and calculated neutron count rates around the primary coolant pipe
Measured count rate (cps) Calculated photoneutron count rate (cps) Delayed neutron count rate calculated by assuming 1.0 · 105 fission in unit coolant volume (cps)
With polyethylene reflector (A)
Without polyethylene reflector (B)
B/A
900 180
280 50
0.31 0.28
620
190
0.31
considering the uncertainties of the above parameters and methods in the calculation and experiment. 4. Sensitivity of fuel failure detection If we regard the calculated photoneutron count rate in Table 1 as the real value at the experiment, we can deduce the delayed neutron count rate in the measurements, and it becomes 720 cps. Since the delayed neutron count rate of 620 cps was obtained by assuming 1.0 · 105 fissions of U235 nuclei in the unit coolant volume of the core, the real number of fission at the experiment can be estimated to be 1.16 · 105 fissions/cm3 s as shown in Table 2. While the fissions occurred in the flow tube, where the nuclear fuels were located, the coolant water going out of the reactor core contained the water through the flow tube and other paths. Since the ratio of the amount of flowing water through the flow tube to that of the core passing coolant is 0.7, the fission rate in the flow tube by surface contamination is calculated to be 1.66 · 105 fissions/cm3 s. In order to estimate the sensitivity of U-235 detection by the delayed neutron measurements, the amount of uranium exposed to the coolant in the core was also predicted. The fission rate of 1 lg U-235 uniformly contaminated on the surface of the fuel rods and exposed to the coolant at the flow tube was calculated by MCNP. The core condition in this calculation was the same as that in the calculation of N-16 generation rate. The calculated value of U-235 fission rate in the flow tube is 8.99 · 107 fissions/lg-U235 s. Using this value Table 2 Deduced delayed neutron count rates and fission rate in the flow tube by surface contamination Delayed neutron count rate (cps) Deduced from measurements and photoneutron calculation (a) Calculated by assuming 1.0 · 105 fissions (b) Fission rate of U-235 in unit coolant volume of the core (1.0 · 105 · a/b) (fissions/cm3 s) Volume ratio of coolant through flow tube to total core outgoing water Deduced fission rate in the flow tube by surface contamination (fissions/cm3 s)
720 620 1.16 · 105
0.7
1.66 · 105
and the estimated fission rate in the flow tube, we can obtain the mass of U-235 which was exposed to the coolant in the flow tube, and it is 1.85 · 103lg/cm3. Considering the total volume of flow tube, the mass of U-235 that was exposed to the coolant in the experiment can be estimated to be 120 lg. The maximum allowable surface contamination of HANARO fuel is 3.25 lg-U235/ft2, i.e., the total allowable mass of contaminated U-235 on the surface of fuel rods loaded in HANARO is 940 lg. Therefore, it is confirmed that the count rate of delayed neutron measured in the primary cooling circuit may have originated from about 13% of the maximum allowable surface contamination. From above results, the sensitivity of U-235 mass detection by delayed neutron measurements can be estimated to be about 0.2 lg-U235/cps. The delayed neutron count rate of about 720 cps at 20 MW reactor power can be obtained from the U-235 exposure of 120 lg, which is much lower than the allowable level of surface contamination of fuel rod. And, the contributions of other components such as photoneutron and gamma-rays are rather small. Therefore, the delayed neutron detection method is sensitive enough to detect the fuel failure, and the neutron count rate is high enough for stable signal with short counting time. In abnormal condition with the fuel failure, the delayed neutron count rates from the failed fuel will be added to the background count rates due to the delayed neutron from the surface contamination and the photoneutron. By using above sensitivity and increased neutron count rate, we can estimate the seriousness of the fuel failure and set the criteria for the safe reactor operation. 5. Conclusion The sensitivity of the fuel failure detection based on the delayed neutron measurement in the primary cooling circuit of HANARO reactor is investigated by both calculation and measurement. The neutron count rate of 900 cps at 20 MW reactor power can be obtained, and about 720 cps is estimated to be the contribution of delayed neutrons. It is confirmed that the delayed neutron count rate can be obtained from the U-235 exposure of 120 lg, which is much lower than the allowable level of surface contamination of fuel rod. The contributions of other components such as photoneutron and gamma-rays are
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rather small. Conclusively, the delayed neutron detection method is sensitive enough to detect fuel failure, and that the neutron count rate is high enough for stable signal with short counting time. References Bak, H.I., Bae, Y.D., Kim, M.S., Choi, H.D., 1995. Peak energy shift dependent on source position in c-ray energy measurement by using a closed-ended coaxial HPGe detector. Nucl. Instrum. Meth. A366, 332– 339. Berger, M.J., Hubbell, J.H., Seltzer, S.M., Chang, J., Coursey, J.S., Sukumar, R., Zucker, D.S., 2005. Downloading Data from INTERNET. Available from: http://www.physics.nist.gov/PhysRefData/ Xcom/Text/XCOM.html. Blatt, J.M., Weisskopf, V.F., 1979. Theoretical Nuclear Physics. SpringerVerlag. Bridwell, L.B., Moak, C.D., 1967. Stopping powers and differential ranges for 79Br and 127I in UF4. Phys. Rev. 156, 242–243. Dobrin, R., Craciunescu, T., Tuturici, I.L., 1997. The analysis of failed nuclear fuel rods by gamma computed tomography. J. Nucl. Mat. 246, 37–42. Gardner, R.P., Barrett, C.L., Haq, W., Peplow, D.E., 1996. Efficient Monte Carlo simulation of 16O neutron activation and 16N decay
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