ÅE-299 UDC 621.039.512.2 621.039.543.4
The Measurement of Epithermal-to-Therma! o
U-238 Neutron Capture Rate (p28) in Ågesta Power Reactor Fuel
G. Bernander
AKTIEBOLAGET ATOMENERGI
STOCKHOLM, SWEDEN 1967
AB-299
THE MEASUREMENT OF EPITHERMAL-TO- THERMAL U-238 NEUTRON CAPTURE RATE (o_) IN ÅGESTA Zo —————— POWER REACTOR FUEL G. Bernander (ASEA)
SUMMARY The epithermal-to-thermal neutron capture rate ratio p_„ in U-238 in Ågesta fuel has been measured by the chemical separation method. The method involves the isolation of Np-239 from uranium and fission products by reversed phase partition chromatography. Although somewhat elaborate, and in spite of difficulties with residual fission products, the method has yielded reasonably accurate results. Further development work on chemical procedures may lead to some improvement. A comparison with the coincidence method - electronic separation of activities - has not shown any large systematic differen ces between the two methods. The separation of the epithermal U-235 activation from the total has been achieved by means of the "l/v subtraction technique" using copper foils as the l/v monitor. The complementary thermal column irradiations required have been performed in the research reactors TRIGA (Helsinki) and Rl (Stockholm). From the measured p_„ values the resonance escape probabi lity (p) and the initial conversion ratio (ICR) may be calculated using cross-section data and other lattice parameters. Comparisons with theoretical values of p and ICR as calculated with the BURNUP lattice parameter code are favourable. The results for the 19-pin cluster of the Ågesta fuel are summarized below.
Tem Experiment Theory pera ture P p ICR P ICR <°0 28
35 0.365 + 0.009 0.902 + 0.003 0.775 + 0.005 0.892 0.795 ( 20°C) 212 0.422 + 0.015 0.882 + 0.004 0.833 + 0.008 0.879 0.838 (220°C)
Printed and distributed in September 1967 LIST OF CONTENTS Page 1. Introduction 3
2. Derivation of p_„ from measured quantities 4 3. Experiments 10 3. 1. Description of fuel and irradiation technique 10 3.2. Measurements 12 3.3. Np-239 activity determination 14 3.4. Auxiliary measurements 19 3.5. Cu foil activity counting 22 4. Results 23 4.1. Activation data 23
4.2. Derivation of p?fi 25 4.3. Calculation of p and ICR 29 5. Summary and discussion 32 Acknowledgements 33 References 34 List of figures 36 - 3 -
1. INTRODUCTION In conjunction with the commissioning physics tests in the Åge sta power reactor [1,2,3] the resonance-to-thermal U-238 neutron capture was rate ratio, p?ft» measured together with several other parameters. This was part of an effort to assess the resonance absorption charac teristics of heavy water moderated and cooled uranium oxide lattices at both ambient and operational temperatures. These characteristics, as represented by p?o» the initial conversion ratio (ICR), or the resonance escape probability (p), are important in predicting initial excess reac tivity and fuel burnup. A chemical method was adopted to separate the fission products from the induced Np-239 activities that were taken as a measure of the neutron capture rate in U-238. Thomasen and Windsor [4] reported favourable results with a chemical separation technique, and the high fluxes available in the Ågesta reactor were an advantage with this method since the high specific activities possible would eliminate the uranium background problem. The "l/v subtraction method", originally proposed by Egiazarov et al. [5], was used to separate the epithermally induced Np-239 activity from the total. In this method the l/v activation is monitored with a separate l/v detector, for which copper foils were employed. In order to obtain absolute values of capture rate ratios, an auxiliary irradiation must be carried out in an essentially pure thermal flux. The fuel samples were 10 mm long UO~ pellets with the same diameter as the ordinary fuel. After irradiation, the pellets were dissolved and a small fraction of the solution was processed to sepa rate the neptunium from the uranium and fission products. The neptunium finally collected and to be counted with an Nal(Tl) crystal was thus in a dissolved state. There are some advantages in using large fuel samples and subsequent dissolution instead of counting very thin foils of fuel mate rial. No irradiation geometry difficulties arise since moderate sur face defects and slight misalignments or gaps between pellets introduce only very small errors. The dissolution yields an averaged and uniform activity distribution in the counting samples, reducing also counting - 4 -
geometry problems. The relatively high irradiation flux necessary in the Ågesta irradiations (because of other simultaneous measurements) also favoured the dissolution technique, which allows an arbitrary activity fraction in a sample to be counted. The high specific activity at the same time reduces problems of natural uranium activity background. A disadvantage of chemical separation work is that exceptionally great care must be exercised in order to avoid excessive activity losses in the various processing stages. The procedures were also rather elaborate and time consuming, especially since the separation yield had not been quite complete, necessitating further complementary measure ments to establish correction factors. In one of the three measurements reported here, a comparison was made with coincidence counting on unseparated dissolved samples. In this case no significant systematic differences between the two methods could be inferred. In the following, the theoretical background is first treated (section 2); then the experiment and the procedures are described (section 3). Section 4 gives the experimental results, and a summary and discussion concludes the paper.
2. DERIVATION OF p2g FROM MEASURED QUANTITIES
In order to emphasize the importance of p?o, the expressions for resonance escape probability (p) and initial conversion ratio (ICR) as a function of p?„ will first be developed; then p?o will be defined in terms of measured quantities.
The definitions of p and p?8 found in the literature vary somewhat depending on the lattice parameter model and the form of cross-section convention used. In this paper we shall conform to the Swedish lattice parameter recipe as represented by the BURNUP code [6] which employs Westcott-type cross-section data. Thus p, as calculated from resonance integral data, involves escape from U-238 resonance absorp tion only. Fast neutron capture in U-238 (above fission threshold) is wholly accounted for in the fast fission factor, whereas all l/v absorption, including that in the resonance region, is contained in the thermal utilization factor. - 5 -
The following notations will be used for the quantities occurring in the formulae: N = neutron radiative capture in U-238 M = neutron absorption (including fission) in U-235 Indices t, r and s denote "thermal" (all l/v capture, re sonance, and fast absorption respectively P , P and P are the fast, intermediate, and thermal non- s r At leakage probabilities f = thermal utilization factor T| = number of fast neutrons produced from U-235 fission per thermal absorption in fuel Q = fast fission factor v = number of fast neutrons formed per fission £ = macroscopic absorption cross-section Indices a, c and f denote total absorption, radiative capture, and fission respectively L, = fast neutron group migration length B2 = the buckling of a critical system F = fission rate a = capture-to-fission cross-section ratio R = total-to-thermal neutron capture ratio in U-238 C = neutron capture in Cu-63
Resonance escape probability (p)
The p-factor is defined as the fraction of the total number of fast neutrons available for slowing down past the U-238 fission threshold that escape capture in the U-238 resonances (excluding the l/v con tribution) during moderation. Then the fraction absorbed in the U-238 resonances is approximately given by (all U-235 capture is accounted for as thermal)
N /P 1 P= " (Nt+M)/PsPrPtpf W
Since the effective multiplication constant may be written as 6 -
k ef,f, = k t» • P srP Pt = Ti-] e rp f • P Psr P.t and
Nt + M = T-(28T 'Nt 3. t
(where £ (U) = £ , (28) + £ (25)) eq. (1) takes the form cl ctX 3.
k S 28 N eff at< > r (?)
In an infinite system k , /P = k , whereas in a finite just ' etir' s <» critical system k ,, = 1 and P = l/(l + L B ).
Now, p?o is defined by
N - N N + N t _ r s ,,, p28 _ N N V> so that
N N N^ = p28 " N^* (4) where the term N /N accounts for the fast capture in U-238 included in p?Q« If the U-235 and U-238 total fission rates are F?(- and F_„ respectively, then
S_(28) N = —^7 r • F = (v F s S. (28) *28 ff28 28 f s and
at N = t ' Sf(25) ' * 25
Combining these two quantities one obtains: - 7 -
Ng £f(25) F^g K = "28 *a7^ ' F25 U
Thus, if F /F is measured or is calculated from the fast fission ' no28' 12c5 factor, if known, according to
F28 , .. V28"1"Q?28 (e-1) F25 V25 then the fast capture contribution may be estimated. Combining eq. (2), (4) and (5) and using the relation
S (25) f Tl ^W V25 one gets instead of eq. (2):
kSsf f Sat(28) keffa28 F28 p = p 6 TIP1:0 * "OuTjatvwr / * 2"8" - I^TP" v25-" s • F"2- 5 < >
The last term in (6) corrects for the fast capture included in the measurement of p?o« This relation is equivalent to the one derived for p by Kouts and Sher [7].
Initial conversion ratio (ICR)
The ICR is defined as the number of fissionable nuclei formed from neutron capture in fertile material per fissionable nucleus con sumed in fresh fuel. Hence, for uranium fuel (neglecting capture in U-234)
T/--D _ Nonfission capture rate in U-238 _ N Total capture rate in U-235 M or N Nt Nt ICR = N;-u = (1+p28)-iM or - 8 -
ICR = (l+p28). f^j (7) av
This relation will be used for calculating ICR from (l + p~R) which is measured. For comparison with theory the "modified relative con version ratio" (RCR*) is frequently used:
* _
where the U-235 fission rate F?c. is determined from a measurement of the fission product activity. The index "th" implies a pure thermal flux. Then the ICR is obtained from
fi The RCR may be expressed also in terms of P-?Q« From eq. (7) and (8) it is easily shown that
Derivation of p^q from measured quantities
p?o has already been defined in eq. (3) as the ratio between re sonance (excluding l/v) plus fast capture and the total l/v capture in U-238. This definition is a consequence of using Westcott cross- section formalism in lattice parameter calculations. Therefore, the cadmium ratio concept is not introduced. One has then:
N + N N - N r s t _N_ . p28 = N " N. = N. t t t
The quantity directly measured is
R = N1 = 1 + p28 <10) - 9 -
For the purpose of subtracting the l/v capture (N.) from the total capture obtained when measuring the Np-239 induced activity, an auxiliary l/v detector is required (Cu foils). If the l/v capture in the Cu foils is denoted by C , eq. (10) may be written as
N S N ^t ,,,, R=S'*TS *? (u) where N/C is the capture rate ratio in the lattice and C /N. = C*/N* may be determined from a thermal column irradiation (asterisks de note thermal column values). Eq. (11) does not take into account the fact that copper has a finite resonance integral. Moreover, the thermal column does have an epithermal flux component, although small. Eq. (ll) is then modi fied as follows. Since
C = Ct. + C r = Ct. x(1+ C r' /C.t')
C* = C* + C* = C*(l+C*/C* v ) t r t r' t'
N* = N* + (N*+N*) = N*(l+N*/Nf+N*/Nf) eq. (11) is converted to
N/N* d+Cr/Ct) (l+NyN*+N*/N*) K _ C/C* * (1+C*/C*) (1ZJ where N, N*, C and C* are determined experimentally from the Np-239 and Cu-64 activities. The correction factors are estimated from cross-section data and auxiliary measurements on epithermal- to-thermal flux ratios. The corrections are treated in more detail in section 4. 2. Even though R may be measured accurately, it is seen from eq. (10) that the uncertainty in p_„ may become rather large when p?R is small. For example, if R has an uncertainty of + 1 %, then p_„ will have an error of near _+ 4 % when p_„ is of the order of 0. 35 (as - 10 -
in the case of the cold Ågesta lattice). Thus (l-p) according to eq. (6) will have an error in excess of J; 4 % and, when account is taken of un certainties in the other data, perhaps some + 5 %. However, if p is approximately 0. 9 (*=» 0. 89 in Ågesta), then the error in p will be less than _+ 0.6 %. This would be satisfactory in view of the rather larger uncertainties in theoretical p values.
In the measurement of p?R care must be taken to irradiate the fuel samples in a lattice region where the neutron spectrum is unper turbed by reflectors or control rods. With Ågesta fuel the spacers between fuel bundles give rise to a thermal neutron flux peak and a corresponding depletion in the epithermal flux component. This effect as a will tend to decrease p?o» consequence, too low ICR-values and too high p-values will be obtained in the vicinity of the spacer gaps. In the experiments the irradiated fuel samples have not been nearer to such a gap than 19 cm, which was thought to be adequate in order to avoid these effects.
3. EXPERIMENTS The procedures may be outlined briefly as follows. UO_ pellets and foils were irradiated in the Ågesta reactor and in a thermal column almost simultaneously. After irradiation the samples were transported to the Studsvik laboratories where Cu foil counting was started imme diately. The UOp pellets were processed in the radiochemical labora tory to obtain separated neptunium samples (solutions) for the counting of Np-239 activity by means of gamma-ray spectroscopy. The fission product solutions and separation column fillings were also collected in order to check Np losses.
3.1. Description of fuel and irradiation technique The fuel in the Ågesta lattice is arranged in a 27. 0 cm square
lattice (fig. 1) corresponding to a moderator-to-UO? volume ratio of 15.6 (including coolant). The experimental fuel assembly was irradiated in a central position (C 10 in the notation of fig. 1) where the neutron flux would be essentially unperturbed by the control rods needed to shim -li
the excess reactivity. The axial flux profile is exemplified in fig. 2 as measured with a copper wire positioned on the outside surface of the coolant channel. The spacer gap flux peaking effect is clearly shown. The fuel assemblies each comprise four 19-pin bundles con nected end-to-end and contained in a coolant tube, see fig. 3a. The total active fuel length is 305 cm including the three spacer gaps, which are 4. 5 to 4. 8 cm each depending on the volume of the expansion space in each pin. The UO_ length in each bundle, then, is 72.5 to 72.8 cm. All structural materials are of Zircaloy. The cross-sectional di- mensions are given in fig. 4a. The UO_ density was 10. 6 g/cm . The special experimental fuel assembly was made from an ordi nary assembly by modifying it to include a removable group of six re presentative fuel pins in one of the four sections (second from bottom end). The pins are normally screwed into each other axially, but in the modified section the experimental pins could be detached laterally through a rectangular opening in the coolant tube, see fig. 3b. The curved plate cut out from the tube was fitted with a top and a bottom grid, thus forming a pin holder. When inserted, the holder was secured with two lock bolts. The integrity of this arrangement was sufficient to withstand normal coolant flow conditions at high tempe rature (220 C) and pressure (34 bars), and permitted handling in the normal manner by the loading machine. The pin holder could be re moved using special shields and tools. The special pins containing the experimental fuel pellets and copper foils (together with various other foils for flux distribution and spectrum measurements) were welded tight prior to insertion into the experimental assembly. After irradiation, samples and foils were removed by sawing off the end plugs of the pins. After irradiation there was a certain delay before the pins could be removed from the experimental assembly. With the low temperature irradiation it was necessary to wait for the fission product activity to decay to an acceptable level - a surface dose rate of some 30 r/h with the shielding arrangement"used. With an irradiation level of some 10 n/cm s for 90 minutes this meant a seven hours' waiting time. After a high temperature irradiation the limiting factor was the cooling- down time and fuel discharge time, altogether some 15 hours. - 12 -
In the thermal column irradiations the fuel specimen has to be of limited size since it is a source of fast neutrons. However, using Cu foils as l/v detector, the condition of equal flux distribution in Cu and adjacent fuel pellet sample is fulfilled only with an extended piece of fuel. Even quite thin samples in an isotropic neutron flux experience a flux depression that is difficult to determine with adequate accuracy.
3.2. Measurements In all, three measurements were made according to table 1 below. The second high-temperature measurement was carried out at an average fuel burnup of 0.23 MWd/kg of uranium. However, the experimental assembly was still of fresh fuel.
Table 1. List of measurements and irradiation data
2) Measure Temperature Sample posi- Control rod insertion Distance of Thermal column 1) . ment tion in sample to irradiation in experimental Control rods nearest con fuel assembly trol rod o No. ( C) (C10) (cm) (cm) (cm)
I 35+1 20.9 A, B, C, D22 230 95 TRIGA (institute of C40 91 Technology, A, B, C, D44 0 Helsinki, Finland) A, B, C, D26 0 A62 <*>50
II 212.5+1 18.9 A, B, C, D44 49 116 Rl A, B, C, D26 0 (Stockholm) A62 92
III3) 212+1 35.5 A, C40; B, D04 199 95 Rl (Stockholm) A, B, C, D44 0 A, B, C, D26 0
1) Distance from center of sample pellet to bottom plug in pin (U0 length in each pin is about 72.5 cm) 2) Position designation according to fig. 1. The rod insertion is counted from "fully in" (0 cm); at 300 cm control rods are wholly withdrawn - such rods are not listed in the table,
3) This experiment was carried out at an average burnup of 0.23 MWd/kgU - 13 -
The UO? pellets used in the experiments (sample pellets and adjacent pellets) were selected from a large batch, taking care to avoid specimens with surface defects. All pellets in the batch had had their ends ground perfectly plane and at right angles to the axis (length variation within J; 0.03 mm), and their diameters were 17.00_+0. 01 mm obtained through centerless grinding. By this means any uncertainties depending upon surface defects or gaps between pellets are quite small. By using 10 and 4 mm long sample pellets in the lattice and thermal column respectively, the effect of misalignments should also be negli gible. In the thermal column the smaller size pellet was desired in order to limit rod size and because of the relatively large neutron flux gradient. Of course, in this case with a near pure thermal flux, sur face defects are not as important as in the lattice. The sample geometries are shown in fig. 4 and 5. Altogether five pellets would be irradiated in each measurement; four in the ex perimental fuel assembly - positions 1, 2, 4 and 5 in fig. 4a - and one in the thermal column. In the lattice rods the copper foils were not positioned next to the sample pellets (see fig. 4b), so that resonance flux streaming effects would be avoided. Since the flux gradient was quite small (compare fig. 2) the Cu activity at the center of the pellet could readily be obtained by interpolation. In the thermal column rod, however, the Cu foils were placed next to the sample pellet (fig. 5). Here resonance flux streaming is of little importance, the thermal flux gradient is large, and it is necessary to avoid end effects on the radial flux distribution. The copper foils were of the same diameter, 17.0 mm, as the UO_ pellets and 0. 10 mm thick. They were protected from catching fission products by sandwiching them between thin aluminium or steel foils 0. 05 mm thick. The Åge sta irradiation was made in a flux of the order of 10 n/cm s for 90 minutes. The reactor power was some 10 to 20 kW but, because coolant circulation flow was the same as at full power (65 MW), no significant excess fuel temperature above that of coolant or moderator was expected. In the high temperature irradiations the coolant and moderator were heated externally by en electric heater and could be maintained at a constant temperature (within _+ 1 C). When - 14 -
possible, the power level was recorded automatically by a digital data acquisition system (RAMSES); otherwise the power was read off and re corded manually. The thermal column irradiations were carried out within a few hours of the Ågesta lattice irradiation. A thermal neutron flux as close as possible to that in the lattice was aimed for; however, this was not always successful due to the power level uncertainty in the Ågesta reac tor. In the Rl reactor thermal column, the sample was positioned at 12. 5 cm from the inner column boundary. Since the graphite reflector is 90 cm, a total of 102 cm of graphite separated the sample from the reactor tank. The D-O reflector of the Rl core is 5 to 10 cm. The sample position in the TRIGA reactor thermal column gave about 127 cm of graphite between the sample and the core. Thus the epithermal flux component in this case should not be greater than in the Rl case. The thermal flux gradient in the thermal column sample positions was 4.8 %/cm in TRIGA and 3.1 %/ cm in Rl.
3.3. Np-239 activity determination Chemical separation The method of separation of neptunium from uranium and fission products (F. P. ) will be described briefly. A more detailed review of the chemical work is found in ref. [8] by S.-E. Kroon, who carried out the chemical work and developed the procedures in collaboration with Å. Hultgren, Studsvik. Basically, however, the method is equivalent to that developed at Kjeller, Norway, by Thomasen and Windsor [4], The principle of the method is the so-called reversed-phase par tition chromatography. An extraction agent (TTA) with which Np(lV) forms a strong complex is adsorbed on hydrophobic material (specially treated glass powder) in a column to form a stationary phase. An aqueous solution with moderate acidity containing the Np, U and F. P. is allowed to run through the column. Np is extracted by the stationary phase whereas U and F. P. are washed away. Then Np is eluted with high-concentration acid and collected. - 15 -
The chemical procedures will now be outlined. a) Dissolving of UO^ Each UO- pellet is dissolved in hot cone. HNC" . With the pellet size used this is a matter of about two hours. Thus a primary stock solution of typically 100 ml is obtained, of which only a small fraction is needed for the subsequent stages. From the stock solution may be taken samples for F. P. counting or for Np-239 counting by the coincidence method. b) Conversion to chloride solution From the nitrate solution obtained in a) is taken a 3 % aliquot in a beaker. After careful evaporation to dryness, cone. HC1 is added. HNO, now escapes and, after further evaporation and adding HC1 once more, the solution is transferred to a 25 ml flask. In this chloride solution the Np has a valence of +6. c) Reduction of Np(Vl) to Np(lV)
Hydroxylaminehydrochloride (NH2OH«HCl) is now added and the reduction is promoted by submerging the flask in a boiling water bath for 1 hour. Then the solution is diluted with water to 25 ml thus obtaining a secondary stock solution containing reduced Np ready for separation. d) Separation of Np from U and F. P.
An aliquot (0. 8 ml containing nearly 0.1 % of the original UO? pellet activity) is now allowed to run through the separation column (described below). Np is selectively extracted by the stationary phase so that U and F. P. only are collected in a con
tainer. A "washing solution" (0. 5M HC1 - 0.1M NH2OH- HCl) is also used to remove all the U and F. P. Then the Np is eluted with 6M HCl- 1M HF plus HCl- saturated CJHgOH plus 6M HCl- IM HF and collected in a separate Teflon container for subsequent activity counting. e) The column glass powder filling is also collected (to determine any residual activity) after drying the columns in an oven. The separation column had an ID of 5 mm in the filling section. The extraction agent, TTA (2-thenoyltrifluoroacetone) dissolved in xylene, was absorbed on hydrophobic glass powder (75- 150 mesh), - 16 -
thus forming the stationary phase of the column. The separation capa city is limited with respect to uranium content in the sample to be separated; for these columns the limit quantity amounts to some 30 mg. Accordingly, the specific Np-239 activity must be rather high to obtain good counting statistics. Naturally, the chemical work will involve inevitable losses of activities in the various stages. In one test three aliquot taken from a single primary stock solution were independently processed to form three "identical" secondary stock solutions (reduced Np). From each of these solutions three aliquots were separated and the activity of the nine samples of separated Np obtained was then counted. The result indicated that the reproducibility of the chemical processes was within _+ 1. 0 % as determined from the spread of results. The averages of the three separations corresponding to each of the three reductions agreed within 0.25 % (i.e. well within counting statistics), showing that the part of the chemical work leading to the secondary stock solution is sufficiently accurate. Apparently the variation in separation yield of Np-239 and/or F. P. residue accounts for the major uncertainty con tribution. The quality of the separation process may be checked by auxiliary measurements (see section 3.4). Later, the samples (aliquots and stock solutions) were carefully weighed instead of merely measuring their volumes. This led to some improvement since it was found that weighing was rather more accurate. Whereas the separation yield of Np is 99.5 % or more, the elution of Np from the column is not quite complete, som 2 to 3 % re maining in it. Since this residual activity is variable, the glass powder filling of the column is collected and its activity measured. It has been found, also, that some F. P. follow the Np, notably Te and Zr. To check this contamination the gamma-ray spectra of all Np, F. P. and glass powder samples were recorded in order to assess correction factors (see section 3.4). For each of the 5 irradiated UO~ pellets, only one reduction was carried out. However, corresponding to each pellet 2 or 3 separations were made, giving a total of 12- 15 separations. - 17 -
Np-239 counting The beta decay of Np-239 into Pu-239 mainly involves the emis sion of gamma-rays with energies of 106.4, 228 and 278 keV. The latter two of these are fairly strongly converted, giving rise to the 103. 7 keV (Kaj and 99- 5 keV (KcyJ X-rays in plutonium. These X-rays combine with the 106.4 keV gamma-ray into the 105 keV photo peak observed in the scintillation spectrometry pulse height spectrum of Np-239. Fig. 6 shows the spectrum obtained with a large Nal (Tl) crystal for both separated and unseparated samples and normalized at 105 keV. In order to suppress efficiently any residual F.P. that may- remain after separation, the Np-239 activity is counted in an interval around the 105 keV peak. The F.P. spectrum is demonstrated in fig. 7, in which the most prominent peaks are identified. No uranium activity is noticeable because of the high specific F.P. activity in the sample. The Np-239 activity in the liquid samples was measured with a 1. 75 in. diam. and 2 in. thick Nal (Tl) crystal mounted on an EMI 6097B photomultiplier tube. The counting equipment included a Nuclear Enterprises non-overloading (NE 5202) linear pulse amplifier, a Landis and Gyr single-channel analyzer (1.2 (j,s resolution time), and an ACEC decade scaler (DM160) provided with an electronic timer. The stabilized high voltage source was also included in the scaler unit. Fig. 8 shows the geometrical counting arrangement. The Teflon containers for the Np-solution had been accurately machined; inter comparison of count rates using a Np-23 9 solution did not reveal differences in excess of the counting statistics error, 0.2 %. The container covers were tightly fastened with adhesive tape to prevent the liquid from evaporating. In spite of this, a slight evaporation was observed (by weighing), probably due to escape of the ethyl-alcohol. Fig. 9 depicts the spectrum of an unseparated sample that was used to set the high voltage and amplifier gain of the counting equip ment. The 105 keV photo peak was positioned between 10 and 15 volts (maximum pulse height was 100 V). Great care was taken to adjust the analyser channel position (lower discriminator level) and width. On the one hand, a small width will reject a maximum of residual - 18 -
F.P. - in an unseparated sample the F.P. contribution in one instance varied from 12. 7 % to 7. 3 % according as the channel width was de creased from 50 to 30 keV (fixed channel position at 86 keV). On the other hand, the effect of electronic drift will be more marked using a narrow "window"; count rate also will be relatively small. The proce dure was to select a channel width somewhat larger than the half- maximum 105 keV line width and then measure the count rate as a function of channel position. The count rate then necessarily experi ences a maximum (provided the channel embraces one peak only), yielding a small "plateau" some 0.4 to 0.6 volt wide inside which the count rate varied less than 0. 25 %. The effect of variable channel width on count rate was typically less than 0. 9 % per 0.1 volt; however, the channel width drift apparently lay well within J; 0. 05 V. Electronic stability was checked in the first two measurements with a Th-228 standard, the count rate of which was very sensitive to gain variations, greater than 3 % per 0.1 V. In the last measurement an Am-243 standard (half-life 7950 y) was used. Except for the 75 keV gamma-ray from the alpha decay of Am-243 into Np-239, the spectrum is pure Np-239 for a fresh sample. The count rate sensitivity was about 0. 8 % per 0. 1 V (due to 75 keV peak) whereas channel width variations should be about the same as for the Np-239 samples, In all cases the checks indicated satisfactory stability. The counting procedure aimed at eliminating any short or long term electronic drifts, in the meanwhile accumulating a sufficient number of counts for good statistics. The samples were counted inter mittently over a span of about two days beginning at about 72 hours after irradiation. The repeated counting of each sample also reduced any random error arising from sample positioning in the counter. The count rates obtained were corrected for dead-time, back ground, and decay. The dead-time correction was significant only in the second measurement with a count rate as high as 2700 c/sec; how ever, all samples had quite similar activities, so the dead-time un certainty became negligible when activity ratios were formed. The detector background was always less than 7 c/sec and the background induced by the sample - as measured on a separated sample from un- - 19 -
irradiated uranium - was only about 1 c/sec. The disintegration con- -4 . -1 stant for Np-239 used in the decay correction was 2. 052 • 10 min (half-life 2.346 days). Finally, the Np-239 count rates were corrected for losses to the separation column and for F. P. contamination - these corrections will be treated in section 3.4 - and were normalized to unit weight of ir radiated UO . The count rate was of the order of 25 to 110 c/sec per mg of U02. 3.4. Auxiliary measurements Since the chemical separation yields were found to vary some what a number of additional measurements were made as a check and in order to assess corrections. The following items were checked: - Np loss to separation column, usually of the order of 2 to 3 %. - Np leakage to F. P. fraction. A leakage in excess of 0. 5 % may be detected but was never experienced in these measure ments. - F.P. residues in the Np fractions. Especially Te and Zr tend to follow the Np. - The overall consistency of the chemical separation process. The following measurements were carried out. a) Counting of activity in column filling The dried glass powder filling was collected and counted in the single channel analyzer. The count rate was added to the count rate of the NP sample since neptunium accounted for essentially all the residual activity in the filling. The error introduced through this correction was at the most about 0.1 % stemming mainly from counting geometry differences. b) Activity sum check Corresponding to every single irradiated pellet two or three separations were made. The consistency of these separations was checked by adding the various activities involved, viz. the count rates from the Np, glass powder, and F.P. samples, and comparing the results. Thus, also the F.P. samples were counted in the single- channel analyzer. - 20 -
In general, the agreement between total sums was quite good, the differences between "identical" samples being compatible with the counting statistics. Evidently, the separation generally did not in volve any variable non-accountable losses. The circumstance made possible an estimate of the variation in separation yields - Np leakage to F.P. fraction and/or F. P. contamination in Np fraction. The apparent yield variation in the Np samples is typically about 1 %. However, corrections for variable F. P. residues (essentially Te-132) in the Np samples reduced these apparent variations to less than 0. 5 %. c) Spectrum analyses The gamma ray spectra of all samples - NP, F.P. and glass powder - were recorded with a Nuclear Data 512-channel analyzer using a large crystal. The purpose was to analyse the spectra for the determination of Np losses or F.P. contamination. Pure reference spectra of Np-239 and F.P. were prepared. In the Np-239 case a 0.1 mm thick natural uranium metal foil was ir radiated under cadmium in the Rl lattice, thus suppressing strongly the U-235 fission rate relative to U-238 capture. The foil was then processed in the same way as the UO? pellets to give a separated Np fraction with a very slight F.P. contamination. Fig. 6 shows the spectrum obtained; the fission product Zr-97 is seen to give a measurable contribution even in this case. A F.P. reference sample was obtained by separating the Np from a high enrichment metal piece irradiated in a thermal column. The "pure" F.P. spectrum is shown in fig. 7 (the uranium background activity is very small and is not noticeable). In the corresponding "Np-fraction", Np-239 activity was barely noticeable, but the partial and variable extraction of Te-132 by the separation column was evident. As mentioned before, Np leakage to the F. P. fraction is be lieved to be less than 0. 5 %. This is apparent when comparing the Np-239 and F. P. spectra. Moreover, any leakage is probably con stant to within about _+ 0. 2 %, as may be inferred from the excellent uniformity of F. P. spectra in the vicinity of 105 keV. - 21 -
Examination of the glass powder activity spectra revealed that the activity was essentially Np-239 together with one major F. P. , namely Te-132. A small correction could be estimated to account for the latter contribution. The predominant F. P. are listed below with their main gamma- ray energies (in keV), compare fig. 7:
Mo99 _67_h^ Tc99m J±0J^ Ru99 ±4±
132 Te132 _78_h_^ j.132 2.3 h^ xe 231, 670
143 143 r 33.4 lu Pr * 294
„ 97 17.0 h^ Nb97 Timing MQ97 665> 75Q Zr > Of these the dominant Mo-99 - Tc-99m activities and the smaller Ce-143 activity appeared to be effectively separated (less than about 5 % of their total activities remain with the Np). On the contrary, Te-132 and Zr-97 were partially extracted together with Np in the separation process. The Zr-97 contamination was rather small and essentially constant and was not corrected for; Te-132, however, con tributed a variable degree of contamination, ranging from near zero to more than 50 % of total Te-132 activity for which a correction was necessary. The fractional contamination of Te-132 in the Np-239 samples was determined from the 231 keV peak intensity in the F. P. spectrum. The size of the correction was established by plotting the differential Te contamination of pairs of identical samples versus the count rate difference of the corresponding Np samples. The error from F. P. contamination is largely systematic and is partly cancelled when activity ratios are formed; the remaining standard error con tribution the these ratios will be about _+ 0.2 %. The uncertainty assigned to the activation of each UO_ pellet as determined from the average of two counted Np samples is generally Jh 0.35 % (standard deviation). This includes uncertainties in irradiation geometry, chemical processes, Np yield and counting, except for errors stem ming from F. P. contamination and the geometry uncertainty in glass powder rest activity counting. The latter uncertainties are more - 22 -
readily introduced when activity ratios are formed, since they are large ly systematic in nature and have been estimated to contribute about _+ 0. 25 % in these ratios. The total root mean square error of lattice pellet to thermal column pellet activity ratios will then be typically 0. 55 %.
3.5. Cu foil activity counting
The Cu foils were gamma counted in a double-detector single- channel analyzer with an automatic foil changer. The double-detector arrangement served to reduce foil positioning errors. The Nal(Tl) crystals were 1. 75 inch in diameter by 2 inches high. The gamma-ray spectra of the two detectors were matched using the Cs-137 600 keV line. The pulses from the detectors were added before amplification and were counted in a single scaler. The analyzer was used as a discriminator, the threshold energy being in the "valley" below the 511 keV annihilation peak. The Co foils were counted in several cycles, thus reducing the effects of electronic instability. The drift was checked by counting a Cs-137 standard sample and was found to be quite small; in fact, no significant error from this source has been attributed to the final Cu-64 activity values. The count rates were automatically punched on paper tape for subsequent computer treatment. The final Cu activity data obtained were corrected for background, dead-time, decay and foil weight dif ferences. The overall uncertainty was less than jf 0.2 % standard de viation in individual foil activities; when four foil activities were averaged to give the interpolated activity at the position of a UO_ pellet this uncertainty was reduced only slightly due to the small uncertainty in the neutron flux distribution along the fuel assembly. Cu-activity ratios then were assigned an error amounting to near _+ 0. 3 %. - 23 -
4. RESULTS 4.1. Activation data
Tables 2 and 3 show the relative Np-239 and Cu-64 activity dis tributions respectively in the fuel cluster. The activities are nor malized to the average cluster activity (N and C respectively) as cal culated from
Ä = ~ (A1 + 6A2 + 6A4 + 6A5)
As is expected, the activation distributions are rather more flat in the hot cases (II and III) than in the cold (I). However, the two hot cases do differ somewhat. From table 2 it is clear that the chemical separation and coincidence methods compare very favourably. The coincidence data are those given by H. Pekarek [9]. In table 3 the Cu activities are compared in cases I and II with the data obtained by Johansson and Sund [10] in the very same irra diations but for near mid-position of the fuel test section. Since the data agree very well, it may be concluded that at least the radial thermal neutron flux distribution is unperturbed by the spacer gap. In this connection it may also be remarked that the effect of the spacer gap thermal flux peaking apparently dies away within about 13 cm along the cluster from the gap [10] as determined from the fission product activity in uranium foils. In table 4 the Np-239 and Cu activity distributions are compared. It is seen that the distributions are remarkably similar.
Table 2. Relative total Np-23 9 activity in fuel cluster normalized to the average activity in cluster (N./N).
Fuel pin I (35°C) II (212.5°C) III (212°C) pos. No. ohem.sep. coino.count 1 0.810 + 0.003 0.835 + 0.003 0.842 + 0.003 0.842 + 0.003 2 0.865 + 0.003 0.884 + 0.003 0.890 + 0.003 0.891 + 0.003 5 1.054 + 0.004 1.055 + 0.004 1.051 + 0.004 1.050 + 0.003 4 1.113 + 0.003 1.088 + 0.004 1.086 + 0.004 1.086 + 0.003 - 24 -
Table 3. Relative Cu activation in fuel cluster normalized to the average activity in the cluster (C./C).
Fuel I (35°C) II (212.5°C) III (212°C) pin pos. This work Johansson
Table 4. Comparison between the relative Np-239 and Cu activity distributions in a fuel cluster (N./N)/(C./C ); the error is about ± 0. 55 %.
Fuel pin pos. I (35°C) II (212.5°C) III (212°C)
No.
1 1.047 1.018 1.021 2 1.014 1.000 1.006 5 0.987 0.999 0.997 4 0.997 0.998 0.997
The lattice-to-thermal column activity ratios, N./N and C./C*, were then formed. These ratios, of course, depend on the actual fluxes in the Ågesta core and the thermal column. In case III the results yielded systematically higher activity ratios for Np (by about 1 %) than the coincidence technique. The discrepancy apparently arose in the counting of the thermal column sample; probably it was the result of the compounded error in the two methods. An unfortunate 1 % uncer tainty for the thermal column sample was, moreover, attached to the coincidence method in this very case. For the chemical separation method the ratios N./N'e were associated with an error of 0.5 to 0.6 %. - 25 -
The copper foil activity ratios, C./C*, had an estimated error of about + 0.3 %. Table 5, finally, gives the values of R? = (NJ/N^JAC./C*). The cluster average values have been calculated according to
N. R1 N;
It is seen that the two hot cases (II and III) differ by nearly 2 % , a dis crepancy that arises when the thermal column activities are taken into account. This will be discussed further in section 4.2 in connection with the p?o results.
Table 5. R? = (N./N*)/(C./C*)
Fuel pin I (35°C) II (212.5°C) III (212°C) poa. No. 1 1.380 + 0.008 1.382 + 0.009 1.411 + 0.009 2 1.337 + 0.008 1.357 + 0.008 1.389 + 0.009 5 1.300 + 0.008 1.356 + 0.008 1.378 + 0.009 4 1.314 + 0.008 1.355 + 0.008 1.378 + 0.009 Cluster 1.320 + 0.007 1.357 + 0.007 1.383 + 0.008 ave.
4.2. Derivation of pOQ
Eq. (12) in section 2 may be written as
p28 + 1 = R = KR' where R' is a quantity derived from two activity ratios according to section 4.1, see table 5, and K is a correction factor:
K = K K3/K2 - 26 -
with
K. = 1 + C /C, (resonance absorption in lattice Cu foils) K_ = 1 + C*/C*(resonance absorption in thermal column Cu foils)
N K-6 = 1 + N*/ f + N*/Nf (resonance and fast absorption in thermal r t s t column u»238).
The correction factor K. is obtained from the expressen (Cu is assumed to have a l/v thermal cross-section)
RI K1=l+ P (13)
<*„o\JJ TT T o'/4 T n + 01» where a o = 4.50 —+ 0.15 b is the 2200 m/' s cross-section of Cu [111. RI = 1. 92 _+ 0. 20 b is the resonance integral (excluding l/v) for a 0.10 mm thick foil [12] . I» = 2[E(kT)/E I1'2 a = 0. 90 a with E = 5 E (kT) is the epi- ^ ^ thermal l/v integral. |3 = r/(l - 1.01 r) where r is the Westcott epithermal index. The quantity r as well as the neutron temperature T have been deter mined by Johansson and Sund [10] for each of the pins in the fuel cluster and are listed in table 6 together with the factor K.; the latter is of the order of 1. 02 to 1. 04 with an error of about 0. 35 %, about half of which is due to the error in the Cu resonance integral.
The other two correction factors, K? and K_, pertain to the thermal column activities. K_ contains a term to account for fast neutron absorption, N*/N*, that cannot be neglected although it is rather much smaller than in a complete cluster. Eq. (5) has been used to estimate this ratio. Ref. [13] gives the fast-to-thermal fission ratio, F28/F25, as (1.81 + 0.08) • 10 for a single UOz rod (corrected for fission source neutrons other than from the rod itself). In our case the thermal column sample is only 5. 5 cm long and should give a some what smaller ratio; however, the quoted value will be retained as a - 27 -
first order approximation. Using N*/N* = 0. 0029 + 0. 0005 The remaining terms in K_ and K~ correct for resonance absorp tion in Cu-63 and U-238; the last term in eq. (13) may be used for the calculation. In order to estimate these activation contributions, the gold foil cadmium ratio has been determined in the thermal column of Rl in order to assess 3« The results are given in the following table: Rcd(T.C.) - 1 R P Cd R d(latt.) - 1 Rl lattice (central channel) 3.24 - 0.0420 + 0.0015 Rl thermal column (T. C. ) at 10 cm from inneinne r end without 640 285 1.5 • 10 -4 uranium source Rl T.C. with 15 cm long , U02 rod { The Rl lattice 3 value has been obtained in other measurements [14] and the thermal column 3 values have then been calculated from the cadmium ratios. Using only a 5. 5 cm long UO? rod, the epithermal flux component will be reduced accordingly, as shown. The g determined in this manner is not wholly appropriate when assessing the Cu-63 and U-238 resonance activations, however. The thermal column sample is a localized source yielding a slowing-down spectrum that decreases slower than l/E when E increases. Therefore, absorbers with their main resonance absorption at energies above that of gold will experience an effectively higher 3 value. In spite of this effect, the 3 obtained above has been used since the resulting error will be quite small. - 28 - The (3 value derived above is not consistent with epithermal cap ture rates inferred from experiments by H. Pekarek [9]. The resulting discrepancy is of the order of 2 to 3 % in R or ICR. A further check of the epithermal contribution in a thermal column with a fission source is under way. The epithermal-to-thermal activation ratios for the Rl column are C*/C* = 2 • 10"4 and Nj/N* = (3.1 + 0.8) . 10"3 so that K2 = 1.0002 + 0.0001 K3 = 1 + 0. 0031 + 0. 0029 = 1. 0060 + 0. 0009 K3/K2 = 1.0058 + 0.0010 In the TRIGA thermal column approximately the same |3 values as in the case of Rl should obtain, since the epithermal contribution from the sample fission source dominates over that from the reactor. The correction factor K is reproduced in table 6, and the values of p?R = R - 1 are given in table 7. The difference of 2 % in R com paring the two hot cases (II and III) results in a difference of about 6 % in p?o» A discrepancy of this order could hardly be caused by dif ferential F. P. contamination. The discrepancy is probably a com bination of random errors in the thermal column samples. The effect of epithermal flux depletion induced by the spacer gap in case II is believed to be negligible, as was the effect of the thermal flux peaking effect mentioned before. The influence of about 0.8 MWd/kgU burnup in the adjacent fuel assemblies should also be unnoticeable. Taking the mean of the cluster averages of p_s for cases II and III, one ob tains p28 = 0.422 + 0.015. - 29 - Table 6. Evaluation of the correction factor K (K-/K = 1. 0058 + 0. 0010 for all cases) Temp. Pin T r K . 1 + C /C K K n 1 r t K. 1 3 (°0 pos. <°C) K2 35 1 127 0.065 1.0362 + 0.0043 1.0422 + 0.0046 2 118 0.059 1.0325 + 0.0039 1.0385 + 0.0042 5 96 0.048 1.0258 + 0.0031 1.0317 + 0.0035 4 94 0.046 1.0246 + 0.0030 1.0305 + 0.0034 212 1 405 0.053 1.0378 + 0.0045 1.0438 + 0.0046 2 381 0.050 1.0352 + 0.0042 1.0412 + 0.0043 5 339 0.044 1.0294 + 0.0035 1.0354+0.0036 4 333 0.043 1.0293 + 0.0035 1.0353 + 0.0036 Table 7. o2„ = R - 1 Fuel pin I (35°C) II (212.5°C) III (212°C) pos. No. 1 0.438 + 0.011 0.443 + 0.012 0.473 + 0.012 2 0.388 + 0.010 0.413 + 0.010 0.446 + 0.011 5 0.341 + 0.010 0.404+0.010 0.427+0.010 4 0.354 + 0.010 0.403 + 0.010 0.427 + 0.010 Ave. 0.365 + 0.009 0.409 + 0.009 0,435 + 0.010 4. 3. Calculation of p and ICR The resonance escape probability as expressed in terms of p?„ by eq. (6) with k ,,. = 1 was then determined. The cross-sections and parameters required for calculating (1 - p) are given in table 1 0 at the end of this section. The fast dif- 2 fusion area L required in determining the corresponding non-leakage probability P was calculated for an effective resonance neutron energy - 30 - of 100 eV using the BURNUP code formula [6], A factor of 2 uncer tainty in the energy value corresponds to only a 0.1 % error in (1 - p). 2 For the buckling B the actual core geometrical value was taken. Table 8 gives the values of p obtained. In all cases the contribution of the last term in the equation for (l - p), i.e. the correction for fast neutron absorption already included in the fast fission factor, was 0.0023 + 0.0002. Table 8. Comparison of measured and calculated p values Measurement 1 -p P p p - p (experiment) (experiment) (theory) exp 'th (*) I (35°C) 0.0977 + 0.0025 0.902 + 0.003 0.892 + 1.2 II (212. 5°C) 0.1142 + 0.0026 0.886 + 0.003 0.879 + 0.8 III (212°C) 0.1217+0.0029 0.878 + 0.003 0.879 -0.1 Ave. II & III 0.1180 + 0.0040 0.882 + 0.004 0.879 + 0.4 The errors quoted originate from the uncertainty in p_~ only. The theoretical values of p are taken from ref. [15] and are calculated with the BURNUP code that uses the resonance integral data of Hell- strand [16]. The theory gives slightly lower values than the experi mental results. The temperature dependence tends to be somewhat larger in the experimental case. The initial conversion rate was calculated according to eq. (7) and is shown in table 9. Since the F. P. activity was not measured accurately, no RCR* values have been obtained directly but may be calculated according to eq. (10) by multiplying the ICR by riiiih 1 + *iat. ^äT^Uh' 1 + «th The error in ICR and RCR''C corresponds to that in R = 1 + p->o» since the cross-section uncertainties have not been included. The theore tical ICR has been calculated according to the BURNUP formalism [17], - 31 - Table 9« Conversion ratio results Measurement ICR Relative RCR* ICR (theory) (experiment) ICR Temp. (°C) I (35°C) 0.775 + 0.005 1.000 1.392 + 0.010 20 0.795 II (212. 5°C) 0.825 + 0. 006 1.065 1.490 + 0.010 220 0.838 III (212°C) 0.841 + 0.006 1.085 1.517 + 0.011 Ave. II & III 0.833 + 0.008 1.075 1.504 + 0.015 Table 10. Lattice parameters used in calculating o and ICR 1 Quantity 35°C 212°C 2 2 L (cm ), slowing down area 84.7 114.0 S to 100 eV 2 - 2 B (m ), geometrical buck 2.16 2.16 ling of core P = 1/(1 + L2 B2) 0.9820 + 0.0010 0. 9760 + 0.0015 s ' x s ' 11 [15] 1.3115 1.2928 e [15] 1.0268 1.0276 V25 2.43 2.43 <*28 0.107 0.107 F /F (lattice cluster [13] 0.052 + 0.0015 0.053 + 0.0015 S L aat(28) (b) 2.698 2.710 aat(U) (b) 7.450 7.336 aat(25) (b) 4.752 4.626 28 S U 0.3621 ^at( )/ at( ) 0.3694 28 25 0.5678 0.5858 Sat< >^a< > 1.0067 1.0120 ^^sW^+^sWcol. - 32 - 5 . SUMMARY AND DISCUSSION 1 The epithermal-to-thermal neutron capture rate ratio (p?o) i* U-238 in the Ågesta reactor fuel has been measured using a chemical method to separate Np-23 9 activity (being proportional to U-238 capture) from uranium and fission products (F. P. ). In chemical work of this type, activity losses are liable to occur; nevertheless, such losses have been reduced to an acceptable level. Rather, the failure of the chemical processes chosen to yield a complete separation sets the limit on ob tainable accuracy. A number of auxiliary measurements were required to determine activity losses and residues in order to assess corrections. These circumstances rendered the measurements and the data analyses somewhat cumbersome. It might also be argued that systematic errors may easily arise. However, a comparison of Np-239 activity ratios with results from the coincidence method (measuring Np-239 in liquid samples) has not revealed any large differences between the two methods. The incompleteness of the chemical separation showed up in the residual Np in the separation column (2 to 3 %) after separation and in the F. P. , notably tellurium and zirconium, remaining with the Np (up to 2 % of total activity). Np "leakage" to the F. P. fraction was less than 0. 5 % and probably varied less than 0. 2 %. These data are the results of studies of gamma-ray spectra obtained with a large detector crystal. A rather small crystal was used to measure the Np-239 activity. Thus, the potential of reducing the F.P. contamination by means of pulse height discrimination has not been fully exploited. A further re lative reduction of the contamination level by 50 % should be possible using a large crystal. More detailed investigations of the F. P re sidues should be made by studying the gamma-ray spectra with a high- resolution lithium-drifted germanium crystal. Any systematic errors arising from incomplete separation yields would in that case be sub stantially reduced. In order to separate the epithermal capture rates from the thermal the "l/v subtraction technique" was employed. The activation of copper foils was used to monitor the l/v capture. A complementary thermal column irradiation is necessary with this method. - 33 - In the measurements p?8 was determined rather than the so called modified relative conversion ratio (RCR*). Either quantity may be used to calculate p and ICR. Whereas p is obtained more directly from p_o than from RCR*, the converse is true for ICR. The difference in the determination of p?fi and RCR*lies mainly in the attainable pre cision with which the l/v capture in U-238 is monitored in the one case (p?o) and the U-235 fission rate in the other (RCR*). The correlation of the two neutron absorption processes - and therefore of p?fi and RCR* - does not involve too great uncertainties. In the former case, the l/v monitor (Cu) is never perfect and a correction for resonance activation has to be applied. When monitoring U-235 fission rate by means of F. P. counting, a correction for U-238 fission rate is neces sary. However, F. P. counting has one specific advantage, viz. the favourable irradiation geometry. No comparison of the two methods of monitoring has been made but is certainly desirable. As a conclusion it may be stated that the counting of Np-239- induced activity by the chemical separation method appears to give reasonably accurate results. The resonance escape probability and the initial conversion ratio as calculated from p?„ agree fairly well with the theoretical values obtained with the BURNUP code employing resonance integral data. ACKNOWLEDGEMENTS The author wishes to thank S. -E. Kroon for his excellent and skilful work in carrying out the neptunium separations and the develop ment work that preceeded it. The author's gratitude is also due to Mr. C.-E. Wikdahl and Mr. H. Pekarek for cooperation and helpful discussions. - 34 - REFERENCES RYDELL, N. , BLOMBERG, P. E. , and ERICSSON, E. , Experience from the commissioning, the criticality experiments and the power operation of the Åge sta Nuclear Power Plant. U.N. Internat. conf. on the peaceful uses of atomic energy, Geneva, 3, 1964. Proc. Vol. 5. New York 1965, p. 421. APELQVIST, G. et al. , Reactor physics studies and comparisons between reactor physics data from calculations and mock-up studies and from measurements in the Åge sta Nuclear Power Plant. U.N. Internat. conf. on the peaceful uses of atomic energy, Geneva, 3, 1964. Proc. Vol. 5. New York 1965, p. 458. McHUGH, B. (ed) , The Åge sta Nuclear Power Station. A staff report by AB Atom energi. Stockholm 1964. THOMASEN, J. and WINDSOR, H. H. , 23g Measurement of resonance absorption in U using a chemical separation technique for isolation of Np^39. 1963. (KR-44). EGIAZAROV, V.B., DIKAREV, V.S. and MADEEV, V.G., Measuring the resonance absorption of neutrons in a uranium- graphite lattice. Conf. Acad. Sci. USSR on the peaceful uses of atomic energy, July, 1955 (AEC-tr-2435, P. 1 p. 59-68). AHLSTRÖM, P.-E. , 1961. AB Atomenergi, Sweden (internal report RFR-151). KOUTS, H. and SHER, R. Experimental studies of slightly enriched uranium, water- moderated lattices. P.l. 1957. (BNL-486). KROON, S.-E. , 1965. AB Atomenergi, Sweden (internal report AP-RKP-42) (in Swedish). PEKAREK, H. , AB Atomenergi. Private communication. JOHANSSON, H. and SUND, L. , Fine structure measurements in the Ågesta Reactor R3. 1964. AB Atomenergi, Sweden (internal report FFX-341; R3-354). HUGHES, P.J. and SCHWARTZ, R.E. , Neutron cross-sections. 2 ed. 1958. (BNL-325). BERNANDER, G. , Unpublished. - 35 - NYLUND, O., Measurements of the fast fission factor (e) in UO_-elements. 1961. (AE-40). JOHANSSON, E. , LAMPA, E. and SJÖSTRAND, N.G. , A fast chopper and its use in the measurement of neutron spectr Arkiv för Fysik _18 (I960) 513-531,(AB Atomenergi, Sweden Internal report RFX-43). APELQVIST, G. , Lattice parameters for the Ågesta reactor fuel. 1962. AB Atomenergi, Sweden (Internal report R3-305; RFR-171) (in Swedish). HELLSTRAND, E. , Measurements of the effective resonance integral in uranium metal and oxide in different geometries. J. Appl. Phys. .28 (1957) 1493-1502. APELQVIST, G. , State Power Board. Private communication. - 36 - LIST OF FIGURES 1 The Åge sta reactor lattice 2 Axial Cu activation distribution along fuel assembly 3 Photograph of experimental fuel assembly 4 Experimental fuel assembly. Irradiation geometry 5 Thermal column sample 6 Np-239 pulse height spectrum 7 Fission product spectrum 8 Np-239 counting geometry 9 Spectrum of unseparated sample Fig, 1 and 2 • Fuel assembly O Control rod po sition Fig. 1. The Ägesta reactor lattice. Lattice positions are identified quadrant-wise using quad rant notation (A toD) and coordinates (without sign). c l Position of test sample Fuel assembly: CIO Temperature: 212.5 °C \- ,0, C g > o Ö 5- o 200 250 300 31 Axial position-cm from bottom plate Fig.2. Axial Cu activation distribution. Report RFT-142. J a) Fig. 3 . Ägesta reactor fuel assembly a) Top section of ordinary assembly with coolant tube cut open b) Special experimental assembly showing demountable set of six fuel pins from an intermediate section. Fig. 4 and 5, /Removable set of six 17.0 'experimental fuel pins. u foils Experimental ^pellet Cu foils Ordinary pellets Fig. A. Experimental fuel assembly. a) Cross section. b) Arrangement of U02 pellets and Cu foils. The foils were shielded against fission products by Al or Fe foils. (All measures in millimetres.) 5 5.0 Pellet lengths 22.7 20 45 2.0 22.7 ^ — É ' f> Direction towards reactor. 17.0 0^ Experimental Cu foils. pellet. Fig. 5. Thermal column sample. The sample was clad in Zircaloy tubing. (All measures in millimetres.) Fig. 6. Np239 pulse height spectrum. Nal(Tl) crystal : Diam. 3 in., height 1,5 in. Np239 (separated sample) - reference spectrum Np239 + U + F. P. (unseparated ) Energy values refer to Np" only. Counting geometry approx. 40 % 2-105- 0) c c o SL o S.105- JO "c zs V-escape o o peak 78 506 F. P. contamination (Zr97) —1 1- 50 100 150 200 250 Energy (Channel no.) Fig. 7, Fission product spectrum. (reference) Mo"-^Tc99m Nal(Tl) crystal ; Diam. 3 in., height 1,5 in. K1 kev Counting geometry approx. 40 % Zr9^Nb97 665 -r— 50 100 150 200 250 Energy (Channel no.) Fig. 8 and 3 W/////////A Teflon container • Np solution Piexiglas holder Nal(Tl) crystal (1,75 in. x 2 in.) Photomultiplier tube (EMI 6097 B) Fig. 8. .239 counting geometry. 105 keV Pulse height (volts) Fig. 9 . Spectrum of unseparated sample with the counting arrangement of fig. 9 LIST OF PUBLISHED AE-REPORTS 266. The half life of the 53 keV level in '"Pt. By S. G. Malmskog. 1987. 10 p. Sw. cr. 10:-. 1-220. (See the back cover earlier reports.) 267. Burn-up determination by high resolution gamma spectrometry: Axial and diametral scanning experiments. By R. S. Forsyth, W. H. Blackladder and 221. Swedish work on brittle-fracture problems in nuclear reactor pressure N. Ronqvist. 1967. 18 p. Sw. cr. 10:-. vessels. By M. Grounes. 1966. 34 p. Sw. cr. 8:-. 268. On the properties of the s, 1, >. a u transition in 1"Au. By A. Bäcklin 222. Total cross-sections of U, UO: and ThOi for thermal and subthermal 3 neutrons. By S. F. Beshai. 1988. 14 p. Sw. cr. 8:-. and S. G. Malmskog. 1987. 23 p. Sw. cr. 10:-. 223. Neutron scattering in hydrogenous moderators, studied by the time de 269. Experimental equipment for physics studies in the Ågesta reactor. By G. pendent reaction rate method. By L. G. Larsson, E. Möller and S. N. Bernander, P. E. Blomberg and P.-O. Dubois. 1967. 35 p. Sw. cr. 10:-. Purohit. 1986. 26 p. Sw. cr. 8:-. 270. An optical model study of neutrons elastically scattered by iron, nickel, 224. Calcium and strontium in Swedish waters and fish, and accumulation of cobalt, copper, and indium in the energy region 1.5 to 7.0 MeV. By B. strontium-90. By P-O. Agnedal. 1966. 34 p. Sw. cr. 8:-. Hoimqvist and T. Wiedling. 1967. 20 p. Sw. cr. 10:-. 225. The radioactive waste management of Studsvik. By R. Hedlund and A. 271. Improvement of reactor fuel element heat trans'er by surface roughness. Lindskog. 1968. 14 p. Sw. cr. 8:-. By B. Kjellström and A. E. Larsson. 1937. 94 p. Sw. cr. 10:-. 228. Theoretical time dependent thermal neutron spectra and reaction rates 272. Burn-up determination by high resolution gamma spectrometry: Fission pro in H:0 and DiO. By S. N. Purohit. 1966. 62 p. Sw. cr. 8:-. duct migration studies. By R. S. Forsyth, W. H. Blackadder and N. Ron qvist. 1967. 19 p. Sw. cr. 10:-. 227. Integral transport theory in one dimensional geometries. By I. Carlvik. 1986. 65 p. Sw. cr. 8:-. 273. Monoenergetic critical parameters and decay constants for small spheres 228. Integral parameters of the generalized frequency spectra of moderators. and thin slabs. By I. Carlvik. 24 p. Sw. cr. 10:-. By S. N. Purohit. 1968. 27 p. Sw. cr. 8:-. 274. Scattering of neutrons by an anharmonic crystal. By T. Högberg, L. Bohlin 229. Reaction rate distributions and ratios in FRO assemblies 1, 2 and 3. By and I. Ebbsjö. 1967. 38 p. Sw. cr. 10:-. T. L. Andersson. 1965. 50 p. Sw. cr. 8:-. 275. ThelAKI=1, E1 transitions in odd-A isotopes of Tb and Eu. By S. G. Malm- 230. Different activation techniques for the study of epithermal spectra, app skog, A. Marelius and S. Wahlbom. 1987. 24 p. Sw. cr. 10:-. lied to heavy water lattices of varying fuel-to-moderator ratio. By E. K. 276. A burnout correlation for flow of boiling water in vertical rod bundles. By Sokolowski. 1966. 34 p. Sw. cr. 8:-. Kurt M. Becker. 1937. 102 p. Sw. cr. 10:-. 231. Calibration of the failed-fuel-element detection systems in the Agosia 277. Epithermal and thermal spectrum indices in heavy water lattices. By E. K. reactor. By O. Strindehag. 1966. 52 p. Sw. cr. 8:-. Sokolowski and A. Jonsson. 1967. 44 p. Sw. cr. 10:-. 232. Progress report 1965. Nuclear chemistry. Ed. by G. Carleson. 1936. 23 p. 278. On the d5 2<-"^97i2 transitions in odd mass Pm nuclei. By A. Bäcklin and Sw. cr. 8:-. S. G. Malmskog. 1967. 14 p. Sw. cr. 10:-. 233. A summary report on assembly 3 of FRO. By T. L. Andersson, B. Brun- 279. Calculations of neutron flux distributions by means of integral transport felter, P. F. Cecchi, E. Hellstrand, J. Kockum, S-O. Londen and L. I. methods. By I. Carlvik. 1967. 94 p. Sw. cr. 10:-. Tirén. 1966. 34 p. Sw. cr. 8:-. 234. Recipient capacity of Tvären, a Baltic Bay. By P.-O. Agnedal and S. O. W. 280. On the magnetic properties of the K = 1 rotational band in "»Re. By S. G. Bergström. 1966. 21 p. Sw. cr. 8:-. Malmskog and M. Höjeberg. 1967. 18 p. Sw. cr. 10:-. 235. Optimal linear filters for pulse height measurements in the presence of 281. Collision probabilities for finite cylinders and cuboids. By I. Carlvik. 1967. noise. By K. Nygaard. 1966. 16 p. Sw. cr. 8:-. 28 p. Sw. cr. 10:-. 236. DETEC, a subprogram for simulation of the fast-neutron detection pro 282. Polarized elastic fast-neutron scattering of "C in the lower MeV-range. cess in a hydro-carbonous plastic scintillator. By B. Gustafsson and O. I. Experimental part. By O. Aspelund. 1987. 50 p. Sw. cr. 10:-. Aspelund. 1966. 26 p. Sw. cr. 8:-. 283. Progress report 1966. Nuclear chemistry. 1967. 26 p. Sw. cr. 10:-. 237. Microanalys of fluorine contamination and its depth distribution in zircaloy 284. Finite-geometry and polarized multiple-scattering corrections of experi by the use of a charged particle nuclear reaction. By E. Möller and N. mental fast-neutron polarization data by means of Monte Carlo methods. Starfelt. 1968. 15 p. Sw. cr. 8:-. By O. Aspelund and B. Gustafsson. 1987. 60 p. Sw. cr. 10:-. 238. Void measurements in the regions of sub-cooled and low-quality boiling. 285. Power disturbances close to hydrodynamic instability in natural circulation P. 1. By S. Z. Rouhani. 1968. 47 p. Sw. cr. 8:-. two-phase flow. By R. P. Mathisen and O. Eklind. 1987. 34 p. Sw. cr. 10:-. 239. Void measurements in the regions of sub-cooled and low-quality boiling. 288. Calculation of steam volume fraction in subcooled boiling. By S. Z. Rou P. 2. By S. Z. Rouhani. 1966. 60 p. Sw. cr. 8:-. hani. 1967. 26. p. Sw. cr. 10:-. 240. Possible odd parity in "sXe. By L. Broman and S. G. Malmskog. 1986. 287. Absolute E1, A K = O transition rates in odd-mass Pm and Eu-isotopes. 10 p. Sw. cr. 8:-. By S. G. Malmskog. 1967. 33 p. Sw. cr. 10:-. 241. Bum-up determination by high resolution gamma spectrometry: spectra 288. Irradiation effects in Fortiweld steel containing different boron isotopes. from slightly-irradiated uranium and plutonium between 400—830 keV. By By M. Grounes. 1967. 21 p. Sw. cr. 10: . R. S. Forsyth and N. Ronqvist. 1986. 22 p. Sw. cr. 8:-. 289. Measurements of the reactivity properties of the Ågesta nuclear power 242. Half life measurements in 1"Gd. By S. G. Malmskog. 1966. 10 p. Sw. reactor at zero power. By G. Bernander. 1967. 43 p. Sw. cr. 10:-. cr. 8:-. 290. Determination of mercury in aqueous samples by means of neutron activa 243. On shear stress distributions for flow in smooth or partially rough annuli. tion analysis with an account of flux disturbances. By D. Brune and K. Jir- By B. Kjellström and S. Hedberg. 1966. 66 p. Sw. cr. 8:-. low. 1967. 15 p. Sw. cr. 10:-. 244. Physics experiments at the Agesta power station. By G. Apelqvist, P.-Å. 291. Separtaion of 5'Cr by means of the Szilard-Chalmers effect from potassium Bliselius, P. E. Blomberg, E. Jonsson and F. Akerhielm. 1968. 30 p. Sw. chromate irradiated at low temperature. By D. Brune. 1967. 15 p. Sw. cr. 8:-. cr. 10:-. 245. Intercrystalline stress corrosion cracking of inconel 600 inspection tubes in 292. Total and differential efficiencies for a circular detector viewing a circu the Ågesta reactor. By B. Grönwall, L. Ljungberg, W. Hiibner and W. lar radiator of finite thickness. By A. Lauber and B. Tollander. 1987. 45 p. Stuart. 19S6. 28 p. Sw. cr. 8:-. Sw. cr. 10:-. 246. Operating experience at the Ågesta nuclear power station. By S. Sand 293. Absolute M1 and E2 transition probabilities in "5U. By S. G. Malmskog and ström. 1986. 113 p. Sw. cr. 8:-. M. Höjeberg. 1967. 37 p. Sw. cr. 10:-. 247. Neutron-activation analysis of biological material with high radiation levels. By K. Samsahl. 1986. 15 p. Sw. cr. 8:-. 294. Cerenkov detectors for fission product monitoring in reactor coolant water. By O. Strindehag. 1967. 58 p. Sw. cr. 10:-. 248. One-group perturbation theory applied to measurements with void. By R. 1e3 Persson. 1966. 19 p. Sw. cr. 8:-. 295. RPC calculations for K-forbidden transitions in W. Evidence for large inerfial parameter connected with high-lying rotational bands. By S. G. 249. Optimal linear filters. 2. Pulse time measurements in the presence of Malmskog and S. Wahlbom. 1967. 25 p. Sw. cr. 10:-. noise. By K. Nygaard. 1966. 9 p. Sw. cr. 8:-. 296. An investigation of trace elements in marine and lacustrine deposits by 250. The interaction between control rods as estimated by second-order one- means of a neutron activation method. By O. Landström, K. Samsahl and group perturbation theory. By R. Persson. 1966. 42 p. Sw. cr. 8:—. C-G. Wenner. 1967. 20 p. Sw. cr. 10:-. 251. Absolute transition probabilities from the 453.1 keV level in 183W. By S. G. 297. Natural circulation with boiling. By R. P. Mathisen. 1967. 58 p. Sw. cr. 10:-. Malmskog. 1966. 12 p. Sw. cr. 8:-. 298. Irradiation effects at 160—240°C in some Swedish pressure vessel steels. 252. Nomogram for determining shield thickness for point and line sources of By M. Grounes, H. P. Myers and N-E. Hannerz. 1967. 38 p. Sw. cr. 10:-. gamma rays. By C. Jönemalm and K. Malén. 1968. 33 p. Sw. cr. 8:-. 299. The measurement of epithermal-to-thermal U-238 neutron capture rate (p28) 253. Report on the personnel dosimetry at AB Atomenergi during 1965. By K. A. Edwardsson. 1966. 13 p. Sw. cr. 8:-. in Ågesta power reactor fuel. By G. Bernander. 1967. 42 p. Sw. cr. 10:—. 254. Buckling measurements up to 250°C on lattices of Ågesta clusters and on D:0 alone in the pressurized exponential assembly TZ. By R. Persson, A. J. W. Andersson and C-E. Wikdahl. 1966. 56 p. Sw. cr. 8:-. 255. Decontamination experiments on intact pig skin contaminated with beta- gamma-emitting nuclides. By K. A. Edwardsson, S. Hagsgård and Å. Swens- son. 1966. 35 p. Sw. cr. 8:-. 256. Pertubation method of analysis applied to substitution measurements of buckling. By R. Persson. 1966. 57 p. Sw. cr. 8:-. 257. The Dancoff correction in square and hexagonal lattices. By I. Carlvik. 1966 35 p. Sw. cr. 8:-. 258. Hall effect influence on a highly conducting fluid. By E. A. Witalis. 1966. 13 p. Sw. cr. 8:-. 259. Analysis of the quasi-elastic scattering of neutrons in hydrogenous liquids. Förteckning över publicerade AES-rapporter By S. N. Purohit. 1966. 26 p. Sw. cr. 8:-. 1. Analys medelst gamma-spektrometri. Av D. Brune. 1961. 10 s. Kr 6:—. 260. High temperature tensile properties of unirradiated and neutron irradiated 2. Bestrålningsförändringar och neutronatmosfär i reaktortrycktankar — några 20Cr-35Ni austenitic steel. By R. B. Roy and B. Solly. 1966. 25 p. Sw. synpunkter. Av M. Grounes. 1962. 33 s. Kr 6:-. cr. 8:-. 3. Studium av sträckgränsen i mjukt stål. Av G. Ostberg och R. Attermo 261. On the attenuation of neutrons and photos in a duct filled with a helical 1963. 17 s. Kr 6:-. plug. By E. Aalto and A. Krell. 1966. 24 p. Sw. cr. 8:-. 4. Teknisk upphandling inom reaktorområdet. Av Erik Jonson. 1963. 64 s. 262. Design and analysis of the power control system of the fast zero energy Kr 8:-. reactor FR-0. By N. J. H. Schuch. 1966. 70 p. Sw. cr. 8:-. 5. Ågesta Kraftvärmeverk. Sammanställning av tekniska data, beskrivningar 263. Possible deformed states in "Mn and "'In. By A. Bäcklin, B. Fogelberg and m. m. för reaktordelen. Av B. Lilliehöök. 1964. 336 s. Kr 15:-. S. G. Malmskog. 1967. 39 p. Sw. cr. 10:-. 1 6. Atomdagen 1965. Sammanställning av föredrag och diskussioner. Av S. 264. Decay of the 16.3 min. "Ta isomer. By M. Höjeberg and S. G. Malmskog. Sandström. 1966. 321 s. Kr 15:-. 1967. 13 p. Sw. cr. 10:-. Additional copies avaiable at the library of AB Atomenergi, Studsvik, Ny 265. Decay properties of 1"Nd. By A. Bäcklin and S. G. Malmskog. 1967. 15 p. köping, Sweden. Micronegatives of the reports are obtainable through Film- Sw. cr. 10:-. produkter, Gamla landsvägen 4, Ektorp, Sweden. EOS-tryckerierna, Stockholm 1967