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Experiences with CR-39 SSNTD in monitoring neutrons emitted from spent reactor fuel
Nucl. Tracks Radiat. Meas., Vol. 19, Nos 1-4, pp. 511-516, 1991 Int. J. Radiat. Appl. lnstrum., Part D Printed in Great Britain
0735-245X/91 $3.00 + .00 Pergamon Press pie
E X P E R I E N C E S WITH CR-39 SSNTD IN M O N I T O R I N G N E U T R O N S E M I T T E D F R O M S P E N T R E A C T O R FUEL
J. P/ILFALVI C e n t r a l R e s e a r c h Institute for P h y s i c s of the H u n g a r i a n A c a d e m y of Sciences H-1525 B u d a p e s t 114, P.O. Box 49, Hungary
ABSTRACT N e u t r o n e m i s s i o n f r o m spent fuel a s s e m b l i e s was s t u d i e d using H u n g a r i a n made CR-39 for m o n i t o r i n g the t r a n s u r a n i u m c o n t e n t as the b u r n u p parameter. To detect t h e r m a l and fast neutrons B o r o n loaded and p o l y e t h y l e n e r a d i a t o r s w e r e applied. It was r e c o g n i z e d that the high gamma b a c k g r o u n g in the spent fuel ponds causes s o m e t i m e s a d r a s t i c change in the b u l k etch rate and in the p a r t i c l e d e t e c t i o n sensitivity. An e m p i r i c a l formula was d e d u c e d to e s t i m a t e the change of the b u l k etch rate:
VBr/VBo =
f ~ t ~ B/T -1/2,
(i)
w h e r e factor f d e p e n d s on the e t c h i n g conditions, t is the e x p o s u r e time, B is the b u r n u p p a r a m e t e r and T is the c o o l i n g time after r e m o v i n g the fuel from the reactor core. In addition, Co-60 and LINAC i r r a d i a t i o n facilities c o m b i n e d w i t h P u - B e n e u t r o n sources were used to study the e f f e c t of the high gamma dose on the bulk etch rate and proton and alpha s e n s i t i v i t y of the H u n g a r i a n CR-39.
KEYWORDS
Spent fuel; burnup;
n e u t r o n detection;
CR-39;
bulk etch rate;
image analyser.
INTRODUCTION The p u r p o s e of the studies - still b e i n g run - was to prove that SSNTDs can be used to i n v e s t i g a t e the b u r n u p of spent reactor fuels and their use can be a c o m p e t i t i v e n o n - d e s t r u c t i v e assay m e t h o d (beside In-ll5 activation, b u b b l e damage polymer, Si diode gross gamma and o t h e r type of d e t e c t o r s ) for s a f e g u a r d i n g spent r e a c t o r fuels. The w o r k has b e e n c a r r i e d out in s t r o n g c o l l a b o r a t i o n w i t h the I n s t i t u t e of Isotopes of the H u n g a r i a n A c a d e m y of S c i e n c e s and the N u c l e a r P o w e r S t a t i o n of Paks and p a r t l y s u p p o r t e d by the IAEA under the R e s e a r c h C o n t r a c t No. 5089/RB. The m e t h o d is b a s e d on d e t e c t i n g neutrons e m i t t e d from the spent fuel. It is w e l l k n o w n that n e u t r o n s can be d e t e c t e d by SSNTDs u t i l i z i n g some type of n u c l e a r reactions as neutron i n d u c e d fission or a l p h a e m i s s i o n or recoil protons. However, in o r d e r to o p t i m i z e the d e t e c t i o n e f f i c i e n c y the n e u t r o n s p e c t r u m and the e n v i r o n m e n t a l c o n d i t i o n s (the gamma c o n t a m i n a t i o n of the n e u t r o n field, the t e m p e r a t u r e etc.) , s h o u l d be known. In the case of s p e n t fuels as n e u t r o n sources (stored under w a t e r ) it can be a s s u m e d that the n e u t r o n s are o r i g i n a t e d from the s p o n t a n e o u s fission of several isotopes of C u r i u m and P l u t o n i u m or from gamma i n d u c e d fission reactions. The e m i t t e d n e u t r o n s are a t t e n u a t e d by the s u r r o u n d i n g water, w h i c h means that the n e u t r o n s p e c t r u m on the p l a c e of the d e t e c t o r spans o v e r about nine o r d e r of m a g n i t u d e s f r o m the thermal range up to roughly 10 MeV. The gamma c o n t a m i n a t i o n of the d e t e c t o r in the storage (cooling) time. and c o n v e r t e r m a t e r i a l s
and the t e m p e r a t u r e h i g h l y s t o r a g e pool, on the b u r n u p These conditions determine and the way of e v a l u a t i o n , 5]]
depends on the l o c a t i o n of the fuel and on the the type of the d e t e c t o r as well. P r e l i m i n a r y
512
J. P/~,LFALVI
studies performed in the spent fuel storage pool of the CRIP's Research Reactor showed that Cellulose Nitrate or Acetate materials cannot stand the temperature of about 60 oC coupled with high gamma dose and the Polycarbonates have not enough intrinsic neutron sensitivity. Therefore, the Hungarian made CR-39 (type MA-ND/p) detector material with 1 m m thick polyethylene and with BNI type Boron loaded radiator (Kodak-PathS, France) was used to detect the fast and the thermal neutrons, respectively. An image analyser system (IMANAL-3) was chosen for the experimental works. The resolution of the system (512 x 512 pixels, 64 gray levels, max. O.3 ~un per p i x e ~ was found to be adequate to measure the necessary track parameters (diameter) to establish and to follow the change of the response functions, which is calculable by the code VIFAN (P~ifalvi, 1991). EXPERImeNTAL WORK Previous experiences showed that the detectors should carefully be prepared and several tests carried out before performing the irradiations in the storage pool of the spent fuels or at calibration sources. The following tests of the sliced detectors were performed: the effect of the cutting process, the smoothness of the surface, the track density and diameter distribution of the background and the uniformity of the bulk etch rate. The batch proved to be the best was selected. To measure the thermal neutrons lOB loaded radiators (BNI) were applied utilizing the l O B ( n , ~ ) 7Li reaction. The ~ particles emitted were detected by the CR-39 material. However, the fast neutron induced recoil protons (from the radiator and also from the upper layer of the detector material itself) produce tracks during the etching process which are mixed up with the ~ tracks. Therefore, a study had to be carried out how to separate the proton and alpha tracks. For the purpose the diameter distribution function of the perpendicularly incident tracks was studied on etched detectors irradiated by neutrons from a Pu-Be source. The results are illustrated on Fig. i. An unfolding procedure (similar to that one used in gamma spectrometry) was elaborated to obtaine the number of the thermal and fast neutrons from the diameter distribution.
=1
CR-39, potyefhy[ene radiotor
20J¢: t.J 0 ,4,,=.
L. 0
I
r'1 I'~
Ill
I--'--"II
CR-39, BN1 rndintor L
E Z
= 20
5
Fig.
10 15 20 Diameter of tracks,
25
30
pm
i. The distribution of tracks by diameter. The detectors were exposed by Pu-Be neutrons moderated by a polyethylene sphere. The etching was made in 6N NaOH at 70 oc for 6 hours.
The same detector preparation process was followed as At was summarized earlier (P~lfalvl et al., 1988). For all the investigations, the detectors were etched in 6N NaOH solution at 70 Oc for different time durations.
N E U T R O N S EMITTED F R O M SPENT REACTOR FUEL
513
First e x p e r i m e n t at the p o w e r s t a t i o n The e x p e r i m e n t a l a r r a n g e m e n t for i r r a d i a t i o n s in the central c h a n n e l of the fuel a s s e m b l y is p r e s e n t e d in Fig 2, w h i l e the c r o s s - s e c t i o n of the capsule w h i c h holds the d e t e c t o r is given in Fig. 3. The d e t a i l e d d e s c r i p t i o n of the w h o l e e x p e r i m e n t a l s e t - u p has b e e n e a r l i e r p u b l i s h e d (Lakosl et al., 1989). Before the actual i r r a d i a t i o n s in the fuel pond, c o n t r o l d e t e c t o r s were irrad i a t e d by alphas from an A m - 2 4 1 s t a n d a r d source to check the f o r m a t i o n of the alpha tracks, also d e t e c t o r s w e r e i r r a d i a t e d by Cf-252 fission fragments to m e a s u r e the r e m o v e d layer thickness in o r d e r to i n v e s t i g a t e the p o s s i b l e c h a n g e m e n t of the bulk etch rate of the d e t e c t o r m a t e r i a l and to e s t a b l i s h the p r o t o n track d e t e c t i o n e f f i c i e n c y (which depends on the r e m o v e d layer). The control d e t e c t o r s w e r e i r r a d i a t e d t o g e t h e r w i t h the virgin ones in the fuel pond. All the detectors (including the c o n t r o l ones and those left to i n v e s t i g a t e the increase of the b a c k g r o u n d ) were e t c h e d t o g e t h e r for a short p e r i o d (2 hours). Then, the n e c e s s a r y param e t e r s w e r e m e a s u r e d by the video a n a l y s e r and e v a l u a t e d by the VIFAN Guide tube code. The results showed some i n c r e a s e of the bulk and the alpha track etch rate, as well. No p r o t o n tracks a p p e a r e d yet. F r o m Steel wire these data the m i n i m u m e t c h i n g time n e e d e d to d e v e l o p fully the alpha tracks was determined. A c c o r d i n g l y , the e t c h i n g was c o n t i n u e d for more 2 hours f o l l o w e d by the e v a l u a t i o n process. It was e v i d e n t that the alpha tracks w e r e w e l l d e v e l o p e d but not the Pond woter p r o t o n tracks, a l t h o u g h their d i a m e t e r was already measurable. For several d e t e c t o r s the track d i a m e t e r d i s t r i b u tions w e r e m e a s u r e d by the video analyser. An ~ t t e m p t was made to esti- Fuel assembly mate the thermal and fast n e u t r o n fluences in this stage of the process Centro[ hole ( L a k o s i et al., 1989). The change of the detector p a r a m e t e r s b e c a m e Sompte holder e v i d e n t and some of the detectors were damaged UO;~ and not e v a l u a b l e further on. F r o m the results it 'ock eh:h defector s e e m e d that the statistical a c c u r a c y of the measIndium sample urements can still be imp r o v e d for the alpha tracks if the total etchFue[ rods ing time reaches the 6 h o u r s and for the proton tracks a l t o g e t h e r 9 hours are needed to have w e l l s h a p e d tracks and reasonFig. 2. E x p e r i m e n t a l a r r a n g e m e n t for ably high d e t e c t i o n efi r r a d i a t i o n s in the central ficiency, so the e t c h i n g channel of the fuel assemblies. was c o n t i n u e d ( i n c l u d i n g The sample h o l d e r is d e t a i l e d the control and b a c k in the next figure. g r o u n d d e t e c t o r s ) and all m
-.....
514
J. PALFALVI
Indium foil
16 Steer
//~
the p o s s i b l e track parameters were m e a s u r e d and e v a l u a t e d again. Finally, it was p o s s i b l e to e v a l u a t e the change of the bulk etch rate as an e f f e c t of the high dose gamma exposure. The e q u a t i o n (I) was d e d u c e d for the bulk etch rate ratio ( V B ~ / V B o ) . The results are s u m m a r i z e d in Table i. The a~ial distzib u t i o n of the change of the bulk etch rate a l o n g & fuel assembly iS given in Fig. 4. For comparison, also the axial d i s t r i b u t i o n of the hard gamma r a d i a t i o n w a s plotted. The gamma doses w e r e d e t e r m i n e d by Si diodes (Lakosi et al., 1989, 1991).
Further studies On the e f f e c t of gamma dose
,
/
track etch detector CR-39 poLyethyLene radiators teflon gasket 7.5
Fig.
3. C r o s s - s e c t i o n a l view of the sample h o l d e r capsule used for i r r a d i a t i n g the n e u t r o n detectors.
To make it clear how the d e t e c t o r s b e h a v e if exp o s e d by high gamma dose, detectors w e r e i r r a d i a t e d by C o - 6 0 gammas of 1.3 t ~ V and by " B r e m s s t r a h l u n g " with m a x i m u m e n e r g y of about 4 ~ V p r o d u c e d by a LINAC. The same e v a l u a t i o n p r o c e d u r e was followed as above. Some of the i r r a d i a t e d detectors w e r e a d d i t i o n a l l y i r r a d i a t e d by e i t h e r a l p h a p a r t i c l e s from an Am-241, by Cf-252 fission fragments or by recoil p r o t o n s p r o d u c e d in a 1 m m thick p o l y e t h y l e n e r a d i a t o r exp o s e d by n e u t r o n s e m i t t e d from a Pu-Be source. The change of the track detection e f f i c i e n c y was s t u d i e d and s u m m a r i z e d in Table 2. No c h a n g e m e n t for the fission fragments was observed.
S e c o n d e x p e r i m e n t at the p o w e r s t a t i o n U t i l i z i n g the results of the p r e v i o u s e x p e r i m e n t s new i r r a d i a t i o n s w e r e performed at the p o w e r s t a t i o n w i t h s h o r t e r i r r a d i a t i o n times to reduce the gamma dose b e l o w 2 kGy and by this way to f a c i l i t a t e the d e t e r m i n a t i o n of the axial d i s t r i b u t i o n of the t h e r m a l and fast n e u t r o n emissions. For i l l u s t r a tions, see Fig. 5.
NEUTRONS EMITTED FROM SPENT REACTOR FUEL
Table i.
Assembly No.
Data of several fuel assemblies and irradiation parameters at the Nuclear Power Station of Paks. (See Eq. l)
Cooling time, T [month]
13673 13616 2908 6672 3409
U-235 cont., [%]
Burnup, B [ M w d / k g ~
3.6 3.6 2.4 2.4 1.6
31.99 30.29 20.41 22.91 11.19
1.6 1.6 15.2 15.1 41.9
Assembly No.
515
Irrad.
13673 13616 13616 13616 2908 6672 3409
time,
t [h]
t B TI/2
9 5 5 1 23.5 23 15
VBT/VBo
f
Gamma dose [kGy]
5.64 3.16 3.02 1.30 3.21 3.30 2.41
0.0262 0.0264 0.0252 O.0542 x 0.0261 0.0244 O.O931 x
30.6 17.0 17.O 3.4 14.O 16.0 not known
215.5 119.8 119.8 24.0 123.0 135.2 25.9
averaged value for 19 detectors
..................... 0.0261 .................. +O.O012
Xnot included in the average
2 1.8 1.6 1.4 1.2
E = 3.6 %
8 2O &.
I
I"
t,J
0.8 0.6 0.4 0.2
o
0
T
3,. 2-5 I
I O
I Fig.
10 9
T = 49 days
I
I
I
2
3
4
I
I
!
I
l
!
5 6 7 Position
8
9 10
4. Axial distribution of gamma radiation inside the fuel assembly No. 13673. Values given by error bars were obtained by 415In~ intensity measurements. Circles represent the measurements by CR-39 bulk etch rate ratio measurements (Vs~/VBo). E is the original U-235 content, T is the cooling-time.
Table 2. The change of the bulk etch rate and particle detection efficiency as the effect of the gamma radiation. Detector: CR-39 MA-ND/p type. Source and dose LINAC, 2 kGy LINAC, 5 kGy LINAC, 10 kGy CO-60, 2 kGy CO-60, 5 kGy CO-60, iO kGy unirradiated NT ]9."U4-~
VBT/VB° 1.O5 + 1.21 ~ 1.40 ~ 1.O7 ~ 1.27 ~ 1.65 + 1 VBo 1 w
O.12 0.07 0.20 O.ii O.iO 0.35 = i. 25 + O. iO ~m/h
Am-241 alphas 1.01 1.O8 0.98 1.04 0.96 0.95 1
Recoil protons by Pu-Be neutrons 0.97 O.91 0.66 0.98 0.86 0.58 1
516
J. P,~LFALVI
E=3.6 %
T=&75days -q
~ , ,
s-~
- 35 ~,
-30
IO0
0eut n -I
cIJ
.m
e=33
zs
&-
QJ
50
-
-
15
-
10
t.J
cl ¢-
I--
10
I
I
I
I
1
I
I
I
I
I
1
2
3
/+
5
6
7
8
9
10
Position Fig.
5. Axial distribution of thermal and fast neutrons inside two fuel assemblies. Detector: CR-39 MA-ND/p type. Min. iOO fields were observed. Statistical error: less than 5%.
CONCLUSIONS The CR-39 track etch detector - even without radiator - can be used to estimate the burnup of spent reactor fuels if the associated gamma dose is higher than 2 kGy. It can be performed by measuring the change of the bulk etch rate and using the expression (i). The factor f is constant above 5 kGy (roughly up to 30 kGy), the dependence on the gamma dose between 2 and 5 kGy is still to be investigated. The change in the bulk etch rate gives a possibility to estimate also the axial gamma dose distribution along the fuel assembly. However, the establishment of the proper conversion factors is necessary. Above 30 kGy gamma dose, the detector surface is completely destroied and the bulk etch rate can be determined only by weight measurement and with a high uncertainty (30-40 %). The bulk etch rate above 2 kGy changes even if the detector is irradiated merely by gammas. For instance, for iO kGy Co-60 gamma dose the increase is approximately 65 %. If during the irradiation the gamma dose remains below 2 kGy then, the surface of the detector is not damaged and the contour of the track is well followable by an image analyser. By applying appropriate radiators an approximate neutron spectrum and fluence can be determined. The axial distribution of neutrons along the fuel assembly can be estimated, as well. Practically, the particle detection efficiencies do not change. It was observed the detection sensitivity for protons decreased if the detectors were irradiated by gammas only with a dose higher than 5 kGy prior to the neutron irradiation. For alphas with energies less than 2 MeV (down to 400 keV) there was no remarkable difference found.
REFERENCES P~ifalvi, J. (1991). A computer code (VIFAN) to calculate the sensitivity of SSNTDs using measured track parameters. To be published in Nucl. Tracks. Lakosi, L., Alm~si, I., Pavlicsek, I., S~f~r, J., FUI6p, L., and Veres, A. (1989). Development of NDA Methods for Power Reactor Spent Fuel Assemblies Using in Activation and Solid State Detectors. Proc. ESERDA llth Symp. on Safeguards and Nuclear Material Menagement, Luxemburg, 1989, Editor: L. Stanchi, Joint Research Centre, Ispra, Italy, pp. 425-430. P~ifalvi, J., Lancsarics, G., Feh~r, I. and S~gi, L. (1988). Alpha Spectrum of "Hot Particles" determined by CR-39 SSNTD. Nucl. Tracks Radiat. Meas., 15 pp. 779-782. La~-6si, L., S~f~r, J., Ffil6p, L. and Pavlicsek, I. (1991). Experience with Bubble Detectors in Monitoring Neutron Emission from Spent Reactor Fuel, this conference.
0735-245X/91 $3.00 + .00 Pergamon Press pie
E X P E R I E N C E S WITH CR-39 SSNTD IN M O N I T O R I N G N E U T R O N S E M I T T E D F R O M S P E N T R E A C T O R FUEL
J. P/ILFALVI C e n t r a l R e s e a r c h Institute for P h y s i c s of the H u n g a r i a n A c a d e m y of Sciences H-1525 B u d a p e s t 114, P.O. Box 49, Hungary
ABSTRACT N e u t r o n e m i s s i o n f r o m spent fuel a s s e m b l i e s was s t u d i e d using H u n g a r i a n made CR-39 for m o n i t o r i n g the t r a n s u r a n i u m c o n t e n t as the b u r n u p parameter. To detect t h e r m a l and fast neutrons B o r o n loaded and p o l y e t h y l e n e r a d i a t o r s w e r e applied. It was r e c o g n i z e d that the high gamma b a c k g r o u n g in the spent fuel ponds causes s o m e t i m e s a d r a s t i c change in the b u l k etch rate and in the p a r t i c l e d e t e c t i o n sensitivity. An e m p i r i c a l formula was d e d u c e d to e s t i m a t e the change of the b u l k etch rate:
VBr/VBo =
f ~ t ~ B/T -1/2,
(i)
w h e r e factor f d e p e n d s on the e t c h i n g conditions, t is the e x p o s u r e time, B is the b u r n u p p a r a m e t e r and T is the c o o l i n g time after r e m o v i n g the fuel from the reactor core. In addition, Co-60 and LINAC i r r a d i a t i o n facilities c o m b i n e d w i t h P u - B e n e u t r o n sources were used to study the e f f e c t of the high gamma dose on the bulk etch rate and proton and alpha s e n s i t i v i t y of the H u n g a r i a n CR-39.
KEYWORDS
Spent fuel; burnup;
n e u t r o n detection;
CR-39;
bulk etch rate;
image analyser.
INTRODUCTION The p u r p o s e of the studies - still b e i n g run - was to prove that SSNTDs can be used to i n v e s t i g a t e the b u r n u p of spent reactor fuels and their use can be a c o m p e t i t i v e n o n - d e s t r u c t i v e assay m e t h o d (beside In-ll5 activation, b u b b l e damage polymer, Si diode gross gamma and o t h e r type of d e t e c t o r s ) for s a f e g u a r d i n g spent r e a c t o r fuels. The w o r k has b e e n c a r r i e d out in s t r o n g c o l l a b o r a t i o n w i t h the I n s t i t u t e of Isotopes of the H u n g a r i a n A c a d e m y of S c i e n c e s and the N u c l e a r P o w e r S t a t i o n of Paks and p a r t l y s u p p o r t e d by the IAEA under the R e s e a r c h C o n t r a c t No. 5089/RB. The m e t h o d is b a s e d on d e t e c t i n g neutrons e m i t t e d from the spent fuel. It is w e l l k n o w n that n e u t r o n s can be d e t e c t e d by SSNTDs u t i l i z i n g some type of n u c l e a r reactions as neutron i n d u c e d fission or a l p h a e m i s s i o n or recoil protons. However, in o r d e r to o p t i m i z e the d e t e c t i o n e f f i c i e n c y the n e u t r o n s p e c t r u m and the e n v i r o n m e n t a l c o n d i t i o n s (the gamma c o n t a m i n a t i o n of the n e u t r o n field, the t e m p e r a t u r e etc.) , s h o u l d be known. In the case of s p e n t fuels as n e u t r o n sources (stored under w a t e r ) it can be a s s u m e d that the n e u t r o n s are o r i g i n a t e d from the s p o n t a n e o u s fission of several isotopes of C u r i u m and P l u t o n i u m or from gamma i n d u c e d fission reactions. The e m i t t e d n e u t r o n s are a t t e n u a t e d by the s u r r o u n d i n g water, w h i c h means that the n e u t r o n s p e c t r u m on the p l a c e of the d e t e c t o r spans o v e r about nine o r d e r of m a g n i t u d e s f r o m the thermal range up to roughly 10 MeV. The gamma c o n t a m i n a t i o n of the d e t e c t o r in the storage (cooling) time. and c o n v e r t e r m a t e r i a l s
and the t e m p e r a t u r e h i g h l y s t o r a g e pool, on the b u r n u p These conditions determine and the way of e v a l u a t i o n , 5]]
depends on the l o c a t i o n of the fuel and on the the type of the d e t e c t o r as well. P r e l i m i n a r y
512
J. P/~,LFALVI
studies performed in the spent fuel storage pool of the CRIP's Research Reactor showed that Cellulose Nitrate or Acetate materials cannot stand the temperature of about 60 oC coupled with high gamma dose and the Polycarbonates have not enough intrinsic neutron sensitivity. Therefore, the Hungarian made CR-39 (type MA-ND/p) detector material with 1 m m thick polyethylene and with BNI type Boron loaded radiator (Kodak-PathS, France) was used to detect the fast and the thermal neutrons, respectively. An image analyser system (IMANAL-3) was chosen for the experimental works. The resolution of the system (512 x 512 pixels, 64 gray levels, max. O.3 ~un per p i x e ~ was found to be adequate to measure the necessary track parameters (diameter) to establish and to follow the change of the response functions, which is calculable by the code VIFAN (P~ifalvi, 1991). EXPERImeNTAL WORK Previous experiences showed that the detectors should carefully be prepared and several tests carried out before performing the irradiations in the storage pool of the spent fuels or at calibration sources. The following tests of the sliced detectors were performed: the effect of the cutting process, the smoothness of the surface, the track density and diameter distribution of the background and the uniformity of the bulk etch rate. The batch proved to be the best was selected. To measure the thermal neutrons lOB loaded radiators (BNI) were applied utilizing the l O B ( n , ~ ) 7Li reaction. The ~ particles emitted were detected by the CR-39 material. However, the fast neutron induced recoil protons (from the radiator and also from the upper layer of the detector material itself) produce tracks during the etching process which are mixed up with the ~ tracks. Therefore, a study had to be carried out how to separate the proton and alpha tracks. For the purpose the diameter distribution function of the perpendicularly incident tracks was studied on etched detectors irradiated by neutrons from a Pu-Be source. The results are illustrated on Fig. i. An unfolding procedure (similar to that one used in gamma spectrometry) was elaborated to obtaine the number of the thermal and fast neutrons from the diameter distribution.
=1
CR-39, potyefhy[ene radiotor
20J¢: t.J 0 ,4,,=.
L. 0
I
r'1 I'~
Ill
I--'--"II
CR-39, BN1 rndintor L
E Z
= 20
5
Fig.
10 15 20 Diameter of tracks,
25
30
pm
i. The distribution of tracks by diameter. The detectors were exposed by Pu-Be neutrons moderated by a polyethylene sphere. The etching was made in 6N NaOH at 70 oc for 6 hours.
The same detector preparation process was followed as At was summarized earlier (P~lfalvl et al., 1988). For all the investigations, the detectors were etched in 6N NaOH solution at 70 Oc for different time durations.
N E U T R O N S EMITTED F R O M SPENT REACTOR FUEL
513
First e x p e r i m e n t at the p o w e r s t a t i o n The e x p e r i m e n t a l a r r a n g e m e n t for i r r a d i a t i o n s in the central c h a n n e l of the fuel a s s e m b l y is p r e s e n t e d in Fig 2, w h i l e the c r o s s - s e c t i o n of the capsule w h i c h holds the d e t e c t o r is given in Fig. 3. The d e t a i l e d d e s c r i p t i o n of the w h o l e e x p e r i m e n t a l s e t - u p has b e e n e a r l i e r p u b l i s h e d (Lakosl et al., 1989). Before the actual i r r a d i a t i o n s in the fuel pond, c o n t r o l d e t e c t o r s were irrad i a t e d by alphas from an A m - 2 4 1 s t a n d a r d source to check the f o r m a t i o n of the alpha tracks, also d e t e c t o r s w e r e i r r a d i a t e d by Cf-252 fission fragments to m e a s u r e the r e m o v e d layer thickness in o r d e r to i n v e s t i g a t e the p o s s i b l e c h a n g e m e n t of the bulk etch rate of the d e t e c t o r m a t e r i a l and to e s t a b l i s h the p r o t o n track d e t e c t i o n e f f i c i e n c y (which depends on the r e m o v e d layer). The control d e t e c t o r s w e r e i r r a d i a t e d t o g e t h e r w i t h the virgin ones in the fuel pond. All the detectors (including the c o n t r o l ones and those left to i n v e s t i g a t e the increase of the b a c k g r o u n d ) were e t c h e d t o g e t h e r for a short p e r i o d (2 hours). Then, the n e c e s s a r y param e t e r s w e r e m e a s u r e d by the video a n a l y s e r and e v a l u a t e d by the VIFAN Guide tube code. The results showed some i n c r e a s e of the bulk and the alpha track etch rate, as well. No p r o t o n tracks a p p e a r e d yet. F r o m Steel wire these data the m i n i m u m e t c h i n g time n e e d e d to d e v e l o p fully the alpha tracks was determined. A c c o r d i n g l y , the e t c h i n g was c o n t i n u e d for more 2 hours f o l l o w e d by the e v a l u a t i o n process. It was e v i d e n t that the alpha tracks w e r e w e l l d e v e l o p e d but not the Pond woter p r o t o n tracks, a l t h o u g h their d i a m e t e r was already measurable. For several d e t e c t o r s the track d i a m e t e r d i s t r i b u tions w e r e m e a s u r e d by the video analyser. An ~ t t e m p t was made to esti- Fuel assembly mate the thermal and fast n e u t r o n fluences in this stage of the process Centro[ hole ( L a k o s i et al., 1989). The change of the detector p a r a m e t e r s b e c a m e Sompte holder e v i d e n t and some of the detectors were damaged UO;~ and not e v a l u a b l e further on. F r o m the results it 'ock eh:h defector s e e m e d that the statistical a c c u r a c y of the measIndium sample urements can still be imp r o v e d for the alpha tracks if the total etchFue[ rods ing time reaches the 6 h o u r s and for the proton tracks a l t o g e t h e r 9 hours are needed to have w e l l s h a p e d tracks and reasonFig. 2. E x p e r i m e n t a l a r r a n g e m e n t for ably high d e t e c t i o n efi r r a d i a t i o n s in the central ficiency, so the e t c h i n g channel of the fuel assemblies. was c o n t i n u e d ( i n c l u d i n g The sample h o l d e r is d e t a i l e d the control and b a c k in the next figure. g r o u n d d e t e c t o r s ) and all m
-.....
514
J. PALFALVI
Indium foil
16 Steer
//~
the p o s s i b l e track parameters were m e a s u r e d and e v a l u a t e d again. Finally, it was p o s s i b l e to e v a l u a t e the change of the bulk etch rate as an e f f e c t of the high dose gamma exposure. The e q u a t i o n (I) was d e d u c e d for the bulk etch rate ratio ( V B ~ / V B o ) . The results are s u m m a r i z e d in Table i. The a~ial distzib u t i o n of the change of the bulk etch rate a l o n g & fuel assembly iS given in Fig. 4. For comparison, also the axial d i s t r i b u t i o n of the hard gamma r a d i a t i o n w a s plotted. The gamma doses w e r e d e t e r m i n e d by Si diodes (Lakosi et al., 1989, 1991).
Further studies On the e f f e c t of gamma dose
,
/
track etch detector CR-39 poLyethyLene radiators teflon gasket 7.5
Fig.
3. C r o s s - s e c t i o n a l view of the sample h o l d e r capsule used for i r r a d i a t i n g the n e u t r o n detectors.
To make it clear how the d e t e c t o r s b e h a v e if exp o s e d by high gamma dose, detectors w e r e i r r a d i a t e d by C o - 6 0 gammas of 1.3 t ~ V and by " B r e m s s t r a h l u n g " with m a x i m u m e n e r g y of about 4 ~ V p r o d u c e d by a LINAC. The same e v a l u a t i o n p r o c e d u r e was followed as above. Some of the i r r a d i a t e d detectors w e r e a d d i t i o n a l l y i r r a d i a t e d by e i t h e r a l p h a p a r t i c l e s from an Am-241, by Cf-252 fission fragments or by recoil p r o t o n s p r o d u c e d in a 1 m m thick p o l y e t h y l e n e r a d i a t o r exp o s e d by n e u t r o n s e m i t t e d from a Pu-Be source. The change of the track detection e f f i c i e n c y was s t u d i e d and s u m m a r i z e d in Table 2. No c h a n g e m e n t for the fission fragments was observed.
S e c o n d e x p e r i m e n t at the p o w e r s t a t i o n U t i l i z i n g the results of the p r e v i o u s e x p e r i m e n t s new i r r a d i a t i o n s w e r e performed at the p o w e r s t a t i o n w i t h s h o r t e r i r r a d i a t i o n times to reduce the gamma dose b e l o w 2 kGy and by this way to f a c i l i t a t e the d e t e r m i n a t i o n of the axial d i s t r i b u t i o n of the t h e r m a l and fast n e u t r o n emissions. For i l l u s t r a tions, see Fig. 5.
NEUTRONS EMITTED FROM SPENT REACTOR FUEL
Table i.
Assembly No.
Data of several fuel assemblies and irradiation parameters at the Nuclear Power Station of Paks. (See Eq. l)
Cooling time, T [month]
13673 13616 2908 6672 3409
U-235 cont., [%]
Burnup, B [ M w d / k g ~
3.6 3.6 2.4 2.4 1.6
31.99 30.29 20.41 22.91 11.19
1.6 1.6 15.2 15.1 41.9
Assembly No.
515
Irrad.
13673 13616 13616 13616 2908 6672 3409
time,
t [h]
t B TI/2
9 5 5 1 23.5 23 15
VBT/VBo
f
Gamma dose [kGy]
5.64 3.16 3.02 1.30 3.21 3.30 2.41
0.0262 0.0264 0.0252 O.0542 x 0.0261 0.0244 O.O931 x
30.6 17.0 17.O 3.4 14.O 16.0 not known
215.5 119.8 119.8 24.0 123.0 135.2 25.9
averaged value for 19 detectors
..................... 0.0261 .................. +O.O012
Xnot included in the average
2 1.8 1.6 1.4 1.2
E = 3.6 %
8 2O &.
I
I"
t,J
0.8 0.6 0.4 0.2
o
0
T
3,. 2-5 I
I O
I Fig.
10 9
T = 49 days
I
I
I
2
3
4
I
I
!
I
l
!
5 6 7 Position
8
9 10
4. Axial distribution of gamma radiation inside the fuel assembly No. 13673. Values given by error bars were obtained by 415In~ intensity measurements. Circles represent the measurements by CR-39 bulk etch rate ratio measurements (Vs~/VBo). E is the original U-235 content, T is the cooling-time.
Table 2. The change of the bulk etch rate and particle detection efficiency as the effect of the gamma radiation. Detector: CR-39 MA-ND/p type. Source and dose LINAC, 2 kGy LINAC, 5 kGy LINAC, 10 kGy CO-60, 2 kGy CO-60, 5 kGy CO-60, iO kGy unirradiated NT ]9."U4-~
VBT/VB° 1.O5 + 1.21 ~ 1.40 ~ 1.O7 ~ 1.27 ~ 1.65 + 1 VBo 1 w
O.12 0.07 0.20 O.ii O.iO 0.35 = i. 25 + O. iO ~m/h
Am-241 alphas 1.01 1.O8 0.98 1.04 0.96 0.95 1
Recoil protons by Pu-Be neutrons 0.97 O.91 0.66 0.98 0.86 0.58 1
516
J. P,~LFALVI
E=3.6 %
T=&75days -q
~ , ,
s-~
- 35 ~,
-30
IO0
0eut n -I
cIJ
.m
e=33
zs
&-
QJ
50
-
-
15
-
10
t.J
cl ¢-
I--
10
I
I
I
I
1
I
I
I
I
I
1
2
3
/+
5
6
7
8
9
10
Position Fig.
5. Axial distribution of thermal and fast neutrons inside two fuel assemblies. Detector: CR-39 MA-ND/p type. Min. iOO fields were observed. Statistical error: less than 5%.
CONCLUSIONS The CR-39 track etch detector - even without radiator - can be used to estimate the burnup of spent reactor fuels if the associated gamma dose is higher than 2 kGy. It can be performed by measuring the change of the bulk etch rate and using the expression (i). The factor f is constant above 5 kGy (roughly up to 30 kGy), the dependence on the gamma dose between 2 and 5 kGy is still to be investigated. The change in the bulk etch rate gives a possibility to estimate also the axial gamma dose distribution along the fuel assembly. However, the establishment of the proper conversion factors is necessary. Above 30 kGy gamma dose, the detector surface is completely destroied and the bulk etch rate can be determined only by weight measurement and with a high uncertainty (30-40 %). The bulk etch rate above 2 kGy changes even if the detector is irradiated merely by gammas. For instance, for iO kGy Co-60 gamma dose the increase is approximately 65 %. If during the irradiation the gamma dose remains below 2 kGy then, the surface of the detector is not damaged and the contour of the track is well followable by an image analyser. By applying appropriate radiators an approximate neutron spectrum and fluence can be determined. The axial distribution of neutrons along the fuel assembly can be estimated, as well. Practically, the particle detection efficiencies do not change. It was observed the detection sensitivity for protons decreased if the detectors were irradiated by gammas only with a dose higher than 5 kGy prior to the neutron irradiation. For alphas with energies less than 2 MeV (down to 400 keV) there was no remarkable difference found.
REFERENCES P~ifalvi, J. (1991). A computer code (VIFAN) to calculate the sensitivity of SSNTDs using measured track parameters. To be published in Nucl. Tracks. Lakosi, L., Alm~si, I., Pavlicsek, I., S~f~r, J., FUI6p, L., and Veres, A. (1989). Development of NDA Methods for Power Reactor Spent Fuel Assemblies Using in Activation and Solid State Detectors. Proc. ESERDA llth Symp. on Safeguards and Nuclear Material Menagement, Luxemburg, 1989, Editor: L. Stanchi, Joint Research Centre, Ispra, Italy, pp. 425-430. P~ifalvi, J., Lancsarics, G., Feh~r, I. and S~gi, L. (1988). Alpha Spectrum of "Hot Particles" determined by CR-39 SSNTD. Nucl. Tracks Radiat. Meas., 15 pp. 779-782. La~-6si, L., S~f~r, J., Ffil6p, L. and Pavlicsek, I. (1991). Experience with Bubble Detectors in Monitoring Neutron Emission from Spent Reactor Fuel, this conference.