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Comparison of Different Detector Performance for the Assessment of Neutron Personal Equivalent Dose in Nuclear Power Plants M. Ginjaume(1), X. Ortega

(1) (2) (1)

, Mª A. Duch(1), M. Tarés(1), F. Fernández(2)

Instituto de Técnicas Energéticas. Universitat Politècnica de Catalunya Grupo de Física de las Radiaciones. Universitat Autònoma de Barcelona

Contact person: [email protected] INTRODUCTION Neutron contribution to personal dose at nuclear power plants is generally low when compared with the gamma component. However, some specific places within the containment can produce significant neutron doses. Due to difficulties in measuring neutron doses in complex neutron/gamma fields, dose estimate in such a situation has often been inaccurate. In some cases the neutron personal equivalent dose is derived from the ambient dose equivalent measured with an area monitor and, in other cases, an individual personal dosemeter is used. In Spain the Nuclear Safety Council (CSN) recommends the use of neutron survey monitors to measure the ambient dose equivalent rate at the workplace and then the individual dose is calculated by taking into account the length of permanence at the monitored site. In spite of this recommendation, some nuclear power plants prefer to use TLD dosemeters for estimating neutron personal dose equivalent. A project, funded by the CSN with the support of the Spanish Nuclear Power Plant Association (UNESA) has been undertaken to investigate in situ the best practical procedure for personal neutron dose monitoring in nuclear power plants. The measurements were performed by two university teams from the Technical University of Catalonia (UPC) and the Autonomous University of Barcelona (UAB) in collaboration with the radiation protection staff from the Ascó, Cofrentes and Almaraz nuclear power plants. The aim of the study was to investigate and to compare the performance of the most widely used neutron survey instruments in Spain with two available TLD systems and some of the new electronic neutron personal dosemeters at some selected sites of three nuclear power plants. In this paper preliminary main results of the first two measuring surveys, in the Ascó I and the Ascó II power plants, are presented. MATERIAL AND METHODS Workplace sites

The measurements were performed at two 1000 MW PWR power plants in the North-East of Spain, Ascó I and Ascó II. Three sites at Ascó I and four at Ascó II, which presented high neutron dose compared with photon dose, were selected for the study. The two PWR plants are twin plants, and the selected measuring points tried to be as close as possible in the two plants. Figures 1 and 2 illustrate the situation of the selected sites in the plant. Sites 1, 2 and 3 correspond to ground level, at the containment equipment door, the edge of the recharge cavity and next to the reactor vessel head. Site 4 was 7.5 m below ground level, next to one of the primary reactor coolant pumps. Because of accessibility to the measuring point and to reduce the individual dose received during the first experiment, for site 2 the position was chosen slightly differently in the case of Ascó II and Ascó I.

Figure 1: Diagram of reactor at ground 7.6 m below level and indication of measuring points measuring at sites 1, 2 and 3 for Ascó I and II. Instruments

Figure 2: Diagram of reactor ground level and indication of point at site 4 for Ascó II.

Spanish nuclear power plants use a Nardeux Dineutron and/or an AlnorStudsvik 2202D for neutron monitoring. Each of these two types of devices represents approximately half of available neutron instruments. The Alnor-Studsvik 2202D(1) consists of a circular cylindrical polyethylene moderator, 47 cm long with 20 cm diameter, with an inner perforated sleeve of boron-loaded plastic and a central BF3 proportional counter. It weighs approximately 10 kg. The Nardeux Dineutron(1) is made of two moderating spheres 6.2 cm and 10.7 cm in diameter, each containing a

cylindrical pressurised 3He proportional counter. This detector is much lighter than typical moderator-based survey meters, it weighs approximately 3.5 kg, which makes it much more convenient for routine measurements. In general, in Spanish nuclear power plants, the AlnorStudsvik is calibrated for 241Am-Be, while the Nardeux Dineutron is calibrated for 241Am-Be and moderated and unmoderated 252Cf (2). The neutron survey instruments evaluated in the study were the Nardeux Dineutron, the Alnor-Studsvik 2202D and the Berthold LB6411(3), which is a promising newer design that could be of some interest for future applications. The Berthold LB6411 consists of a cylindrical pressurised 3 He proportional counter located in the centre of a moderating 25-cm sphere. It weighs approximately 10 kg. As passive personal dosemeters, the two types of TLD used, respectively, in the Ascó and in the Almaraz Nuclear power plants were included in the study. The Harshaw 8814 dosemeter(4) is a beta-gamma-neutron dosemeter. It consists of three 6Li F chips (TLD600) and one 7LiF chip (TLD700) covered by different filters. The neutron personal dose equivalent is calculated using DOELAP software(4). The second tested TLD is an Alnor albedo-type dosemeter(5), UDEL P1700. It consists of two pairs of 6LiF-7LiF detectors (TLD600-TLD700) on a boron plastic sleeve; one pair is on the outside of a thermal-neutron absorber, while the other is inside. This configuration allows the detection of low-energy albedo neutrons backscattered from the user's body. For TLDs, nuclear power plants usually perform calibration in the field by comparison of TLD readings in an appropriate phantom with ambient dose equivalent measured by a calibrated neutron survey meter. To complete the overview, two neutron electronic personal dosemeters were also verified at some sites: the recently commercialised SAPHYMO Saphydose-n(6,7) and the pre-commercial version of the MGP 2000GN(8) electronic dosemeter. The Saphydose-n uses a large silicon strip detector covered by different converters and absorbers. The MGP 2000GN uses two silicon detectors for the determination of photon and neutron personal dose equivalent. The neutron detector is covered by appropriate converters and absorbers to improve its energy response. For the evaluation of the tested monitors and dosemeters, spectral neutron distributions at the analysed workplaces were measured with a Bonner 11-sphere spectrometer. The spectra are obtained using the spectrum unfolding programme MITOM developed at the UAB(9). Subsequently, the ambient dose equivalent is obtained by folding the energy distribution with the fluence-to-ambient dose equivalent conversion coefficients in ICRP 74(10). The Bonner spectrometer does not provide information on the angular energy distribution, therefore, to estimate the personal dose equivalent rate, some geometrical assumptions will be made based on the response of personal dosemeters situated on different sides of the irradiation phantom.

Passive and electronic personal dosemeters were exposed in a 30cmx30cmx15cm PMMA slab phantom. At each site, 4 passive dosemeters and an active dosemeters were placed in the preferred direction (towards the reactor), one TLD and a second active dosemeter, when available, on the opposite side of the phantom, and two TLDs on the other sides of the phantom. The exposure time varied from 2 h to 14 h to obtain an acceptable response. For the survey meters, the measurements were performed in the integration mode to reduce statistical uncertainty. In this case the exposure time varied from 1 to 30 minutes. Figures 3, 4 and 5 show an example of the measurement set-up for the survey meters, the personal dosemeters and the Bonner Spheres.

Figure 3: Experimental set up: irradiation Figure 4: Experimental set up: irradiation of Dineutron at site 3 of personal dosemeters at site 4

Figure 5: Experimental set up: Bonner sphere measurements at site 2

RESULTS Table 1 summarizes the main dosimetric characteristics of the seven surveyed workplaces: total neutron fluence rate measured with the Bonner spectrometer; spectrum mean neutron energy calculated from the experimental spectra; neutron ambient dose equivalent rate derived from experimental fluence spectra and photon ambient dose equivalent rate measured with TLDs. The uncertainty associated with the neutron ambient dose equivalent rate is about 7 % (1 SD). Nuclear power plant Ascó I

Measuring point site I-1 site I-2 site I-3 site II-1 site II-2 site II-3 site II-4

& n (n/cm2)

E

(keV) 34 46 21 22 17 19 19

& H * (10) n

Bonner

& H * (10)

TLD

(µSv/h) 600±40 1850±150 122±9 470±30 570±35 117±10 167±10

(µSv/h) 87±6 210±13 29±2 77±2 135±20 30±4 300±5

Ascó II

5000±200 12200±500 1380±55 5000±200 7000±300 1310±50 1940±80

Table 1: Main dosimetric characteristics of measuring points Figure 6 shows the energy distribution of neutron fluence rate measured with the Bonner spectrometer at the 7 sites. The fluence distribution corresponds to the characteristic shape of a PWR containment site. Measurements at Ascó I were performed at the end of the fuel cycle, whereas measurements at Ascó II were performed just after reload. This could explain the systematic slight higher energy for the same type of location between spectra at Ascó I and Ascó II, since in the first case there was less Boron available in the reactor for neutron moderation.

1600

1400

site I-1

site II-1

site I-2

site II-2

site I-3

site II-3

site II-4

1200

-2 -1 ·E (cm ·s )

1000

800 II-1 600 I-1 II-2 I-2

400

200 I-3,II-3 0 1,E+00

II-4

1,E+01

2,E+01

3,E+01

4,E+01

5,E+01

6,E+01

7,E+01

8,E+01

9,E+01

1,E+02

E (MeV)

Figure 6: Energy distribution of neutron fluence rate measured at the 7 selected sites of a PWR containment. Figure 7 illustrates the energy ambient dose equivalent distribution calculated from the neutron fluence measurements. It highlights that high energy neutrons are mainly responsible for the neutron ambient dose equivalent.

200 180

I-2

site I-1

160 140

-2 -1 ·E (cm ·s )

site II-1

site I-2

site II-2

site I-3

site II-3

site II-4

120 100 80 60 II-2 40 20 0 1,E+00

I-1 II-1

II-4 I-3,II-3 1,E+01 2,E+01 3,E+01 4,E+01 5,E+01 6,E+01 7,E+01 8,E+01 9,E+01 1,E+02

E (MeV)

Figure 7: Energy ambient dose equivalent distribution calculated from neutron fluence measurements, applying ICRP 74 conversion coefficients. The performance of the three surveyed instruments is summarized in Figure 8. The graph shows the ratio between the corrected instrument reading and the Bonner Sphere measurement at each workplace. The corrected instrument reading is calculated by multiplying the reading by the calibration factor for 241Am-Be. Most instrument responses (16/21) were found to be within ± 30 %, which can be considered as a very good agreement between area monitoring and reference values, for most radiation protection practices. However more detailed remarks can be drawn by looking at the individual results more closely. The calibration certificates for three different Dineutron show that the calibration factor of this type of instrument varies from 0.5 for moderated 252Cf to 1.1 for 241Am-Be. A similar energy response was reported in an international intercomparison of neutron survey

instrument calibration(11). Figure 8 shows that the Dineutron underestimates H*(10) by approximately 20 % to 40 %, depending on the measuring point. These difference has been found even though the 241Am-Be calibration factor was applied, thus confirming the fact that this calibration energy is the best for using this instrument in a nuclear power plant environment. The built-in algorithm used by the Dineutron to correct for the encountered field is not available to the user and is based on ICRP 21 recommendations(12). Therefore, from the experimental results one could propose a field-specific correction factor of 1.5 for PWR reactor fields of mean energy of approximately 20 keV.

1,40

+30%

RESPONSE: H*(10)det/H*(10)Bonner

1,20

1,00

0,80

0,60

-30%

0,40

Nardeux Dineutron Berthold LB 6411 Alnor 2202D + BTC TIN

0,20

0,00 I-1 I-2 I-3 II-1 II-2 II-3 II-4

SITE

Figure 8: Ratio between the corrected survey instrument reading and the Bonner Spectrometer measurement at each workplace. The Berthold and the Alnor-Studvisk monitors do not need field-specific factors, since their response is within +/- 30 % of the reference value except for site I-2. At this measuring point the differences between the reference value and the area monitors are around 40 %. This behaviour could be partially explained by the highest mean energy of this site, these changes in response are consistent with published response characteristics for such instruments (2,13).

3,50

Alnor Udel

3,00

Harshaw 8814 DMC 2000GN Saphydose (AP)

RESPONSE: Hp(10)det/EBonner

2,50

(ROT)

(AP)

2,00

(AP)

1,50

(AP) (AP)

(ROT)

1,00

0,50

Figure 9: Ratio between personal dosemeter reading and the effective dose at each workplace. The results for the personal dosemeters are summarized in Figure 9. Contrary to the case of ambient dose equivalent, a reference value is not available for personal dose equivalent because some knowledge of the field angular distribution is needed to derive it from spectrometry data. Therefore, in this case, the graph shows the ratio between the personal dosemeter reading and the effective dose rate calculated from the Bonner spectrometer fluence distribution and the corresponding conversion coefficients tabulated in ICRP 74. The geometrical assumptions are derived from the readings of personal dosemeters situated on different sides of the irradiation phantom, which seem to indicate that the neutron field can be approximated to an AP-geometry except at sites I-1 and II-1 where a rotgeometry seems more appropriate. The ratio of the personal dosemeter reading and the effective dose rate ranges for most measurements between 1 and 2. The electronic dosemeter Saphydose-n overestimates the effective dose in the two tested points a factor of 2.3 and 3.2, respectively. Results show that except TLDs at site I.2, the rest of measurements give a conservative estimate of the effective dose, as it is desirable from a radiation protection point of view. Although the spread of relative response for personal dosemeters is wider than for area monitors, their performance can be considered satisfactory for routine monitoring, and is consistent with other works(14), as well as with the dependence of the effective dose and the personal dose equivalent for the neutron energy distribution in the experimental measuring points shown in ICRP 74(14). CONCLUSIONS The experimental neutron distribution spectra obtained in this work confirm the well- known characteristic neutron field of a PWR reactor containment and provides precise reference values for ambient dose equivalent rate at the workplace. The comparison of these reference values with the readings of the most commonly used neutron survey instruments has allowed some recommendations on the use and interpretation of these measurements to be established in routine monitoring. The Berthold and Studvisk response is generally better than the Dineutron, although there is

precise information on their energy response, specific field corrections are not needed in normal practice. On the other hand, for the Dineutron, a field-specific correction factor is recommended to improve its accuracy. This device is light and handy to use, but its energy dependence on ambient dose equivalent is higher than with the other devices. In particular, it is shown that, in this case, one should prefer the calibration factor of 241AmBe than moderated 252Cf. The evaluation of personal neutron dosemeters at workplaces requires further work since it has not been possible to gather accurate information on the angular field distribution at the selected workplaces. However, preliminary results show good consistency between the different personal dosimetry techniques and the effective dose. The field calibration procedure used by the nuclear power plants for neutron TLD calibration ensures acceptable results. As regards the new neutron electronic dosemeters, only a few measurements have been performed. Their performance does not yet improve albedo TLD performance, but these instruments are foreseen to be useful tools for neutron operational dosimetry. However, for quantitative measurements, more investigation is recommended since at some of the sites, the dose was significantly overestimated. In conclusion, the study confirms that the approaches to estimating the neutron personal dose equivalent from area monitoring or individual dosimetry are both acceptable and consistent. Survey meters could be considered to be more accurate and less energy dependent than personal dosemeters, but they have the disadvantage of their weight and of the difficulty in exactly reproducing workers' movements. In spite of the bad energy response, albedo TLDs calibrated in the field are recommended for individual monitoring. The tested electronic personal dosemeters have been proven to be interesting complementary tools to passive dosimetry as personnel warning devices, but some improvement are still advisable. Acknowledgements The authors would like to thank Ascó and Cofrentes staff for their help and collaboration during the measurements. Thanks are also due to Ascó and Almaraz dosimetry services for providing TLDs and to the manufacturers of the electronic personal dosemeters Synodys MGP and Saphymo for lending us their instruments for the project. The project has been funded by the Spanish Nuclear Safety Authority, contract reference PR-15, 2004.

References

1. ICRU Report 66 Determination of operational dose equivalent quantities for neutrons, 2001. 2. ISO 8529-1 Reference Neutron Radiations ­ Part 1: Characteristics and Methods of Production, 2001.

3. Klett A. and Burgkhardt B. The New Remcounter LB6411: Measurement of Neutron Ambient Dose Equivalent. H*(10) according to ICRP60 with High Sensitivity, IEEE TRANSACTIONS ON NUCLEAR SCIENCE, VOL. 44, 3, 757-759, 1997. 4. Bicron Technologies INC. STI/Harshaw Manual. Dose Algorithm for the accreditation programm of Laboratories of the Department of Energy. Ref ALGM-D-U-0591-002, 1991. 5. Piesch E. and Burgkhardt B. A Universal Beta/Gamma/Neutron Albedo Dosemeter for Personnel Monitoring . Radiation Protection Dosimetry 6:281283, 1983. 6. Saphydose. Electronic Personal Dosimeter for Neutron. Technical characteristics. In www.saphymo.fr. 7. Lahaye, T., Chau, Q., Ménard, S., Ndontchueng-Moyo, M., Bolognese-Milsztajn , T. and Rannou, A. Numerical and experimental results of the operational neutron dosemeter `Saphydose-N'. Radiat. Prot. Dosim. 110, 1-4, 201-206, 2004. 8. MGP 2000 GN. Personal electronic dosemeter technical characteristics. In www.mgpi.com. 9. Fernández F., Bakali M., Tomás M., Muller H., Pochat J.L. Neutron measurements in the Vandellòs II nuclear power plant with a Bonner Sphere system. Radiat. Prot. Dosim. 110, 1-4, 517-521, 2004. 10. ICRP Publication 74. Conversion coefficients for use in radiological protection against external radiation, 1996.

11. Hunt J.B., Champlong P., Chemtob M., Kluge H., Schwartz. International Intercom-parison of neutron survey instrument calibrations. Radiat. Prot. Dosim. 27, 2, 103-110, 1989. 12. ICRP Publication 21. Dataset for Neutron and Photon, 1971.

13. Bartlett D.T., Tanner R.J., Tagziria H., Thomas D.J. Response characteristics of Neutron Survey Instruments. NRPB-R333(rev) Didcot, 2001. 14. Luszik-Bhadra, M., Bolognese-Milsztajn, T., Bartlett, D., et al. Summary of personal neutron dosemeter results obtained within the Evidos project. European workshop on individual monitoring of ionizing radiation. Abril 2005, Viena (Austria).

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