CORRELATION BETWEEN FUEL FAILURE RATE AND PRESSURE DROP-TIME

As fuel failure rate depends on ∆T [7], the correlation between number of failure fuel assemblies (in % from number of fuel assemblies of certain mode) and product ∆²P⋅τ (where τ - time, ∆²P – pressure drop) was detected (Fig.13). It follows, to achieve the value n%≈0 in the course of 3 days it's necessary to keep up the pressure drop on the level ∆²P<0.03 Bar (pH₃₀₀≈6.95÷7.05) (see Fig.5).

0 20 40 60 80 100 120

0 100 200 300 400 500 600

∆²P*τ, bar*day n%

FIG. 13. Correlation between number (%) of failed FAs and parameter ∆²P⋅τ.

5. CONCLUSIONS

1. The phenomenon of "pressure drop" which takes place in NPP with WWER reactors was considered.

2. A model was developed to explain pressure drop rise in-core and deposits redistribution in the core and in the primary circuit of NPP with WWER-440. The physical-chemical basis of the model is the transport corrosion products dependence on temperature and pH_T of coolant.

3. A model was verified by operation data of 7 units of NPP with WWER-440 reactors. A model can be used for water chemistry control during lifetimes and under planning of decontamination measures.

REFERENCES

[1] TUTNOV, A., ALEXEEV, E., “Analysis of different models of clad-to-coolant transfer coefficient influence on results of WWER fuel element thermo-mechanic behaviour”, Proc.

Fifth International Conference "WWER Fuel Performance, Modelling and Experimental Support". 29 September–3 October 2003, Albena, Bulgaria. ISBN 954-9820-09-2. 2004, p.408–416.

[2] BENEDEK P., LASZLO A. Grundlagen des chemieingenieurwesens. Veb Deutscher Verlag Fǜr Grundstoffindustie, Leipzig, 1967.

[3] OVCHINNIKOV, F.Y., GOLUBEV, L.I., DOBRYNIN, V.D., KLOCHKOV V.I., SEMENOV, V.V., V.M.TSYBENKO, Operational regimes of water-cooled nuclear power plants, Moscow, “Atomizdat” publishing house, 1977, p. 280.

[4] ROSENBERG, R.J., TERÄSVIRTA, R., HALIN, M., SUKSI, S., Investigation of iron deposits on fuel assemblies of the Loviisa 2 WWER-440 reactor, «Water chemistry of nuclear reactor systems 7» Proceedings of the conference organized by the British Nuclear Energy Society and held in Bournemouth on 13–17 October 1996 British Nuclear Energy Society,

London. p.27–33.

[5] KERESZTÚRI, A., BOGATYR, S., MIKÓ, S., NEMES, I., Analyses for licensing of new fuel types at NPP Paks, FIFTH INTERNATIONAL CONFERENCE ON WWER FUEL PERFORMANCE, MODELLING AND EXPERIMENTAL SUPPORT. 29 September–

3 October 2003, Congress Center Albena, Bulgaria, ISBN 954-9820-09-2. 2004, p.243–252.

[6] KOMAROV, L.B., Statistical methods of processing experimental data, Lensovet Leningrad Technological Institute, Leningrad, 1972, p. 207.

[7] KRITSKY, V.G., Influence of water chemistry regimes on fuel cladding failure in LWRs.

Fuel failure in normal operation of water reactors: experience, mechanisms and management.

Proceedings of a Technical Committee Meeting held in Dimitrovgrad, Russian Federation.

26–29 May, 1992. IAEA-TECDOC-709 June, 1993, Vienna, p.282–286.

LOAD WITHSTAND ABILITY OF PHWR FUEL BUNDLE END PLATE P.N. PRASAD, R. SONI

Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushaktinagar, Mumbai, India

Abstract

The Indian 220 MWe PHWRs use Natural Uranium short length Fuel Bundles. The fuel bundle consists of 19 fuel elements, arranged and assembled together to form a fuel bundle assembly, with end plates welded at both ends. The fuel bundles experience variety of loads during its stay in the coolant channel namely external coolant pressure, loads due to fuelling machine operation and differential axial thermal expansion of elements etc. The fuel bundle components such as cladding and endplate have been designed with minimum Zircaloy content, with the result the components go in to the plastic range of the stress-strain curve. These components have been designed for fatigue. Differential fissile content in different rings of fuel bundle will affect the thermal flux profile across the bundle. This leads to different heat generations in different ring elements and different axial expansions. Consequently, the end plate gets stressed and the number of fatigue cycles the end plate and the fuel bundle can take will vary. This paper discusses the loads acting on fuel bundle under different scenarios and end plate stress analysis and fatigue cycle estimation for natural uranium bundle and bundle with different fissile content in different rings.

1. INTRODUCTION

Presently, twelve Pressurized Heavy Water Reactors (PHWRs) and two Boiling Water Reactors (BWRs) are in operation in India with a total installed capacity of 2720 MWe. Construction of the two PHWR 500 MWe units at Tarapur, four additional PHWR 220 units at Kaiga (Kaiga 3, 4) and RAPP 5&6, Two VVER-1000 reactors at Kudankulam (KK-1&2), in collaboration with Russian Federation, are in progress.

PHWRs use 'Natural' uranium in dioxide form as fuel. The fuel is in the form of small cylindrical pellets. These pellets are loaded in Zircaloy-4 (Zirconium-Tin-Fe-Cr alloy) cladding and hermetically sealed at both ends by welding with two end plugs. Each element is provided with bearing pads/spacer pads to provide inter-element spacing. The elements are assembled in the form of a bundle by welding them to two end plates. The fuel bundle is made up of elements arranged in two/three concentric rings around a central element. These bundles are half meter in length. The fuel bundle structural components namely, fuel cladding, end plugs, bearing pads, spacer pads and end plates are all made of zircaloy material. Typical fuel bundle is shown in Figure-1.

These bundles are located horizontally in coolant channel (pressure tube) in the reactor. The fuel bundle generates heat by nuclear fission and this heat is transported to the primary coolant. The fuel bundles are replaced after achieving the design discharge burnup. On power bi-directional fuelling is carried out by sliding the bundles in the channel with the aid of two fuelling machines, one at each end of the coolant channel. Pressure tubes containing a string of short length fuel bundles and the on-power refueling permit flexibility in choosing fuel designs and in-core fuel management parameters to maximize fuel utilization. As a part of nuclear fuel cycle development, in addition to regular use of natural uranium fuel bundles, other materials namely depleted uranium, thorium and MOX bundles have been tried in the reactors.

The fuel bundles are designed to generate heat by nuclear fission at heat ratings and conditions consistent with overall core design of the reactor, and have to pass through any coolant channel of the reactor while being subjected to pressure, temperature and flow conditions of the coolant in the channel. During the residence period of the fuel bundle, the zircaloy structural components have to withstand the loads on them due to fuel thermal expansion and change in pellet shape during irradiation like pellet ridging, fission product swelling etc. The fuel bundles experience variety of loads during its stay in the coolant channel namely external coolant pressure, hydraulic flow loads, loads due to fuelling machine and due to differential expansion of elements etc. They are designed with sufficient strength to allow on-power refuelling and withstand fuel-handling procedures in the reactor. The fuel bundle components such as cladding and endplate have been designed with minimum Zircaloy content, with the result the components go in to the plastic range of the stress-strain curve.

The fuel cladding for PHWRs is of collapsible type i.e. the cladding collapses on the fuel pellet stack due to external coolant pressure. These components have been designed for fatigue. The endplate is a very important component of the PHWR fuel, as it holds all the fuel elements together with all conditions. The ability of the endplate to withstand the element differential expansion loads for bundles with different fuel materials/contents is analyzed in this paper.