EXTENDED RCCA OPERATION AND POST-IRRADIATION EXAMINATION

The results of high flux tests of an RCCA model in a material test reactor and commercial operation experience have created premises for RCCA lifetime extension. In order to provide lifetime extension, the RCCAs of unit-5 of the Novovoronezh NPP had been operated for longer time comparing with the period given by the project. After irradiation, these RCCAs have been examined at the Research Institute of Atomic Reactors (RIAR) and the Novovoronezh NPP hot cells. The research was performed and paid by the Russian and Ukrainian utilities (ROZENERGOATOM and GOSKOMATOM) [1-3].

The value of the neutron flux was checked up by means of ¹⁰B burnup and accumulation of activation products (60Co). The main results are given in Table II [1-3]. Completed tests and further research on CRs from CG have confirmed the expected outcome. Vibre compacted CRs have the best operating performances in comparison with joint drawn CRs.

The divergence in maximum thermal neutron flux (Fmax) in CRs between calculated data (2.1 xlO2 1 by the Kurchatov Institute and OKB GP) based on 10B burnup (4.5x1021) and data based on

60Co accumulation in the steel cladding of the CR (6.0xl0²¹) has been determined through SG CR research. The calculated results appeared to be underestimated.

The analysis of all research, resulted in establishing the values for several parameters, i.e. the maximum allowable ¹⁰B burnup for vibro compacted CRs was 57.5%, the CR swelling (A®) was less than 1.2 % (implying that there were safety and plasticity margins of the cladding material) and the relative thermal neutron flux was 3.6xl0²¹.

TABLE II. RESULTS OF RCCA RESEARCH

material B4C state

Mechanical (Absorber of vibrating packed type) C G 2 y , 491 eff. days, All of CR have saved the form and tightness.

There is black dense oxide film the thickness of which 4-7 microns on a surface

0 max. ~32.5 70-130 [cmj] at 20°C 2-4 bar

at 333°C 4-8 bar Powder B4C freely hailed down under At the CR bottom there was a sintering of a powder, which densely CR are close to extreme allowable condition. CR 4. EXPERIENCE OF RCCA LIFETIME EXTENSION

4.1. Operation time

The operation time of RCCAs for CG was enlarged from 1 to 2 fuel cycles [3].

4.2. Life extension

The unconditional specification of the RCCA lifetime extension for SG from 5 to 6 years (and even to 7 years) was not possible, based on the results of the research, as the calculated thermal neutron flux density above the reactor core was lower than the actual value.

In order to be able to specify the SG RCCA lifetime extension, the exact position of the RCCA lower flat end above the reactor core for each unit in each fuel cycle is needed. The lower flat end of RCCA would be within the 0.9 - 14.9 cm above the core in serial WWERs-1000 (e.g. for unit-5 of Novovoronezh NPP - 7.7-8.7 cm). The thermal neutron flux density within the same range should not be a constant value. The dependence of the allowable time of SG RCCA operation according to the position of the lower flat end is given in Table III [2].

TABLE III. ALLOWABLE OPERATING TIME OF RCCAs Position of the lower flat end

640 above the core

Approximately 10 - 15 % of Ukrainian WWER-1000 SG RCCAs have been operated during the 6 fuel cycles. The total duration of operation did not exceed 1700 -1790 effective days. The experience of SG RCCA lifetime extension to 6 and more years has not received broad attention for the reason of difficulties related to monitoring of the RCCA lower flat end position. During the whole period of operating WWERs-1000 in the Ukraine, there occurred no RCCA damage. The materials used in the RCCAs performed reliably.

5. RESULTS OF IMPROVED RCCAs BASED ON A COMBINATION OF Hf AND B4C

The cost, allowable time of operation, and possibility of inexpensive disposal are the main consumer features of RCCAs. In order to increase the RCCA lifetime, it is required to replace the bottom (300-500 mm) n/oc absorber B4C by an n/y absorber (Hf, Dy2Ti05or In-Ag-Cd, which doenot swell) and to use cladding material that will be more stable to radiation embrittlement.

12 RCCAs with combined Hf-B4C absorber were designed, manufactured and accepted by an interdepartmental commission. In 1997, they were loaded in the WWER-1000 of the Rovno NPP for the operation according to a joint decision of GOSKOMATOM and MPP (with participation of Kurchatov institute, RIAR, OKB GP). The CR cladding was made from Cr-Ni alloy EP-630Y. The design of the combined absorber CR is shown in Figure 2. It was decided to use the new designed RCCA in SG.

The ingots made in the Ukrainian zirconium plant, based on the calcium-thermal recovery technology and further double electronic radial remelting, were used as raw Hf source material.

Further processing and manufacturing of Hf rods for CRs were carried out in Russia. Samples of Hf pipes with a diameter (0) 9.6 mm, wall thickness of 2 mm, length 100 mm (9.6x2 mm) and rods with a diameter of 8.2 mm, length 10 mm were stage by stage irradiated and investigated in RIAR for the period 1993- 1997.

Irradiation of samples located in ampoules were made in the material test reactors (BOR-60, in a sodium environment at a temperature of 340 - 360°C, and in CM-2, in water at 60 - 80°C). At the first stage, the irradiation was performed in BOR-60 by a fast neutron flux of (E> 1 MeV) 1 x 1022.

n

\ 7 rod cladding 8,2 x 0,5 Cr - Ni alloy

V

FIG. 2. Combined absorber rod design (Hf+ B4C)

The diameter of the pipe decreased by 0.3% and that of the rod by 0.8%. The lengths of the pipe and rod were increased by 0.3% and 1%, respectively after the irradiation. The densities of the pipe and rod have decreased by 0.42% and 0.55%, respectively. The pipe has plasticity at the level of 1-2% and there was no damage. Two longitudinal radial cracks, which could have occurred at the rod manufacturing stage, were detected. The Hf rod was damaged (an = 4 kgm/cm2) [4] by shock tests.

For that reason the Hf rod has been located inside the cladding. As cladding material, the Cr-Ni alloy (Cr-42%, Ni-56%, Mo-1%) was used. This material named as EP-630Y has a high plasticity after long irradiation and a high corrosion stability (resistance) in a water environment. At the second stage, the irradiation was performed in BOR-60 and CM-2 by a fast neutron flux of (E>1 MeV) 3.4x10". The average radiation growth of the sample was 3%. After the first stage of irradiation, cracks of the same form and size were detected in the rods [5]. The mechanical properties of EP-630Y are given in Table IV [6].

TABLE IV. MECHANICAL PROPERTIES OF EP-630Y Temperature

[°C]

20

300

Neutron fluencex 1 022

E>8MeV [cm'2]

The behaviour of modelled CR was investigated under accident conditions:

• B4C, Hf, Dy2TiO5 absorbers do not affect the EP-630Y cladding at temperatures between 350 and 500°C during 1000 hours;

• The CR research under simulated cladding leakage conditions (temperature of overheated steam of 1150°C) has shown, that there is reliable compatibility of Hf and Dy2TiO5. The Hf modelled CR surface damage was revealed at a level of 400)am. The melting occurred in the Hf cladding contact area under a vapour temperature of 1250°C [10].