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U-238 Neutron Capture Rate (P28) in Ågesta Power Reactor Fuel

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U-238 Neutron Capture Rate (P28) in Ågesta Power Reactor Fuel 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.
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