Post irradiation examination of the fuel discharged from the Trino Vercellese reactor after the 2nd irradiation cycle

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Scientific and technical research
Target audience: Scientific

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• Publié par: EUROPEAN-COMMISSION
• Nombre de lectures: 58
• Langue: English
• Poids de l'ouvrage: 2 Mo

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COMMISSION OF THE EUROPEAN COMMUNITIES
nuclear science and technology
Post irradiation examination of the fuel
discharged from the Trino Vercellese reactor
after the 2nd irradiation cycle
EUR 5605 e COMMISSION OF THE EUROPEAN COMMUNITIES
nuclear science and technology
Post irradiation examination of the fuel
discharged from the Trino Vercellese reactor
after the 2nd irradiation cycle
P. Barbero, G. Bidoglio, M. Bresesti, R. Chevalier, D. D'Adamo,
S. Facchetti, A. Federici, G. Guzzi, L. Lezzoli, F. Mannone,
F. Marell, P.R. Trincherini
(Chemical Division)
G. Buscaglia, A. Drago, R. Faccelli, A. Frigo, E. Ghezzi, R. Klersy,
K.H. Schrader, A. Schuerenkaemper
(Materials Division)
R. Dierckx (Physics Division)
J. Biteau (ESSOR)
G. Cottone, A. Cricchio, L. Koch
(Transuranium Institute, Karlsruhe)
R. Bannella, M. Paoletti-Gualandi, P. Peroni
(Ente Nazionale per l'Energia Elettrica, Italy)
Joint Research Centre - Ispra Establishment - Italy
Joint Research Centre - Karlsruhet - Germany
Ente Nazionale per l'Energia Elettrica - Italy
1977 EUR 5605 e Published by the
COMMISSION OF THE EUROPEAN COMMUNITIES
Directorate-General
' Scientific and Technical Information
and Information Management '
Bâtiment Jean Monnet - Kirchberg
LUXEMBOURG
© ECSC, EEC, EAEC, Luxembourg 1977
Printed in Belgium
LEGAL NÖTIGE
Neither the Commission of the European Commu­
nities nor any person acting on behalf of the Com­
mission is responsible for the use which might be
made of the following information. TABLE OF CONTENTS
1. Introduction and Summary 5
2. Fuel Characteristics and Selection of Fuel Samples 6
3. Metallurgical Examination
10
3. 1 Optical Inspection 24
3. 2 Metrologicai Examination 15
3. 3 Tensile Test on Cladding Sections6
3.4 Metallographicn
4. Gamma Scanning7
5. Determination of Burn-up and Isotopie Compositions 2 1
5.1 Experimental Analysis1
5. 1. 1 Gamma Spectrometry 2 1
5.1.2 Radiochemical Procedures3
5. 1. 3 Massy8
5. 1. 4 Alpha Spectrometry 34
5. 2 Processing of the Experimental Data6
5. 2. 1 Burn-up Determination from 13^Cs 3
5.2.2pn from 1^^Nd
39
5. 2. 3 Burn-upn from Heavy Isotopes 42
5. 2.4 Determination of Isotopie Composition, Build-up 4
and Depletion of Heavy Isotopes
5. 3 Analysis of the Accuracy of the Values of Burn-up and Iso­
topie Composition 55
5. 3. 1 Comparison Between Burn-up Values Determined by-
56
Different Experimental Techniques
5. 3. 2n Between Values of Isotopie Composition,
57
Build-up and Depletion of Heavy Isotopes Determined
at Ispra and Karlsruhe
5.3.3 Isotope Correlations
59
6. Comparison Between Theoretical and Experimental Data of Iso- / Q
topic Composition
7. Conclusions ,
62
References 63 -δ-
1. INTRODUCTION AND SUMMARY
The pos t­irradiation analysis of the fuel discharged from the Trino
Vercellese reactor after the 2nd irradiation cycle, was started in
the framework of the EURATOM­ENEL research contract No. 071­66­6
TEEI­RD and was completed in the framework of the research pro­
gramme of the Joint Research Centre "Technical Support to Nuclear
Power Stations".
The post­irradiation analysis was carried out in the Ispra and Karls­
ruhe Establishments of the Joint Research Centre.
The main objective of the post­ir radiation examination programme
was the measurement of the burn­up and isotopie composition of selec­
ted fuel samples in order to obtain a set of data to be used for check­
ing the accuracy of nuclear code calculations.
The other objective of the programme was the metallographic analysis
of the U02 fuel and of the stainless steel cladding.
imilar programme had been carried out on the fuel discharged. . A s
m the Trino Vercellese reactor, after the 1st irradiation cycle* . fro
The fuel assembly selected for the analysis (No. 509­069) was disman­
tled at Ispra, in the pool of the ESSOR reactor.
The removed fuel rods were subjected to the following examinations
(in the ADECO and LMA laboratories): optical inspection, metrology,
mechanical tests on the fuel cladding and metallography.
In the ADECO and LMA laboratories gamma scanning measurements
on the fuel rods and gamma spectrometry measurements on selected
rod positions were also carried out.
Fuel samples to be subjected to radiochemical analyses were cut from
the rods at selected positions.
The fuel samples were dissolved in the laboratories of Ispra and Karls­
ruhe and aliquots of the solutions were subjected to radiochemical pro­
cesses and to gamma, mass and alpha­spectrometry determinations.
Gamma spectrometry was used mainly to determine the 13'Cs activity,
from which the burn­up was derived.
Mas s­spectrometry, combined in some cases with isotope dilution
techniques, was used to determine the concentrations and/or isotopie
compositions of the heavy isotopes uranium, plutonium, americium,
of 1¿*8Nd and of the krypton and xenon fission gases.
Alpha­spectrometry was used to determine the concentrations of some
nuclides of plutonium, americium and curium. The s of
the heavy isotopes and 148Nd were then used for separate evaluation
of the burn­up. In section 5.1 all the original data of the gamma, mass and alpha-
determinations are reported. These data were processed in order
to derive values of burn-up and of isotopie composition, build-up
and depletion of heavy isotopes, which are reported in section 5. 2.
In the data processing nuclear data were used and some assump­
tions and approximations were introduced.
The availability of the original experimental data, reported in sec­
tion 5.1, also allows those interested to apply a data processing
based on different nuclear data and different assumptions.
In order to check the accuracy of the values of burn-up and isotopie
composition measured in our experiments three procedures were
applied:
- Use of different methods in the measurement of the same quanti­
ty: this procedure was applied for the burn-up values which were
determined from 137Cs, 148Nd and heavy isotopes.
- Comparison between the results of different laboratories: this pro­
cedure was applied by analyzing 5 pairs of adjacent pellets in the
laboratories of Ispra and Karlsruhe.
- Use of isotope correlation techniques (see section 5. 3).
In section 6 a comparison is presented between the experimental
data and the data calculated by ENEL using a nuclear code.
In this comparison some data determined from the fuel discharged
from the Trino Vercellese reactor after the 1st irradiation cyclei1),
are also included.
2. FUEL CHARACTERISTICS AND SELECTION OF FUEL SAMPLES
The Trino Vercellese Nuclear Power Plant, operated by ENEL,is
equipped with a pressurized water reactor rated at 250 MW(e).
Westinghouse Electric Corporation is the designer and the manufac­
turer of the nuclear steam generating plant, fuel included.
The first irradiation cycle of the Trino Vercellese Nuclear Power
Plant started on the 23rd of October 1964 and finished on the 28th
of April 1967; the second irradiation cycle started on the 20th of
May 1970 and finished on the 9th of July 1971.
Descriptions of the reactor core thermo-hydraulic characteristics
and of the core mechanical data during the 2nd irradiation cycle, are
given in Tables 1 and 2. At the end of the 1stn cycle the
core was reduced by substituting 8 fuel assemblies with 8 dummy as­
semblies.
The fuel assembly No. 509-069 selected for pos t-ir radiation ana­
lyses had an initial enrichment of 3. 13 W% in 235U and reached an TABLE 1 - Core Thermo-hydraulic Characteristics
82 5 MW(th) Power output
140 kg/cm2 Coolant pressure
Coolant effective flow rate 16.000 t/h
269°C Coolant average temperature
Core average power density 69.9 kW/l
(22.8 kW/kgU)
11.4 kW/ft Max. design linear power density TABLE 2 - Core Mechanical Data
CORE
240. 0 cm Equivalent diameter
264.9 cm Active height
112 Number of square fuel assemblies
52 r of cruciform fuel s
3 Number of regions
2.72-3,13-3.90% Initial enrichments (square assemblies)
2.72% l enrichment (cruciform )
4.00% Reload Assemblies enrichment
28 Number of control rods
39.626 kg UOo in square fuel assemblies
2.313 kg UO2 in cruciform fuel s
41.939 kg Total U02 weight
36.968 kg Total U-weight
SQUARE FUEL ASSEMBLY
15 χ 15 Rod array
208 Number of fuel rods
20.00 cm Side of square cross section
320. 88 cm Total length
353.81 kg U02 weight
40 Number of reload assemblies
FUEL PELLET
96. 5% T.D. UOo density-
0. 890 cm Diameter
1. 53 cm Length
0. 33 mm Dishing depth
173 Number of pellets per rod (approx. )
264. 1 cm Length of pellet stack in fuel rod
0. 114 mm Clad-pellet clearance
FUEL CLAD
0. 902 cm Inside diameter
0. 383 mm Wall thickness
SS AISI 304 Mate rial
CRUCIFORM FUEL ASSEMBLY
26 Number of fuel rods
240. 3 c