I 1
Atomic Energy of Canada Limited
EXPERIENCE WITH ZIRCONIUM-ALLOY PRESSURE TUBES
by
P.A. ROSS-ROSS, W. EVANS and W.J. LANGFORD
Presented at the HI Inter-American Conference on Materials Technology Rio de Janeiro, Brazil, August 14-17, 1972
Chalk River, Ontario August 1972
AECL-4262, EXPERIENCE WITH ZIRCONIUM-ALLOY PRESSURE TUBES1
by
P.A. Ross-Ross, VV, Evans and W.J. Langford
ABSTRACT
Pressure tubes, which contain the fuel and coolant, are the pressure vessels of tin- Canadian designed CANDU nuclear reactor. A power reactor basically consists of a large number of identical pressure tuoe assemblies. Zirconium alloys are used as tube material because of their good strength, good corrosion resistance, and low neutron absorption. The advantages of the pressure tube include simplicity of design, ease of fabrication, and development of inspection and testing methods with low investment costs. Because pressure tubes are relatively small, prototype lubes can be irradiated at full power raactor conditions in the test reactors. Their behaviour is monitored periodically and finally they are removed for destructive evaluations. The results from strength, corrosion and other evaluations are applied to the design of tubes for power reactors, and to an assessment of their safety. In-reactor creep tests showed that creep was enhanced by fast neutron flux, but also provided the information the designer needs to accommodate creep deformation. This paper briefly describes the Canadian experience with pressure tubes. Of major importance has been the tube irradiation program, which has provided the information and confidence needed to proceed with an ever expanding power reactor program.
1 Presented at the III Inter-American Conference on Materials Technology, Rio do Janeiro, Brazil, August 14-17, 1972.
Chalk River Nuclear Laboratories Chalk River, Ontario August, 1972 AKCL-1262 Experience acquise avec les tubes de force en alliage de zirconium'
par
P.A. Ross-Ross, W. Evans et W.J. Langford
Resume
Les tubes de force qui contiennent le combustible et le caloporteur sont les cuves sous pression des reacteurs de la filifire CANDU. Le reacteur de puissance de type canadien est constitue, essentiellement, par un grand nombre de tubes de force identiques. Ces tubes sont fabriques en alliages de zirconium parce que ceux-ci om, entre autres, les proprietes suivantes: solidite; bonne resistance a ia corrosion et faible absorption des neutrons. Les tubes de force ont, entre autres, les avantages suivants: simplkite de conception, fabrication aisee, methodes d'essai de l'inspection n'entrafnant pas de grandes depenses en immobilisation. Du fait que les tubes de force sont relativement petits, leurs prototypes peuvent etre irradies, en reacteur d'essai, dans les conditions reelles des reacteurs de puissance. Leur comportement est verifie periodiquement et finalement on leur fait subir des essais destructifs. Les resultats obtenus dans les essais (solidite, corrosion, etc.) sont employes dans la conception c!es tubes destines aux reacteurs de puissance, ce qui permet d'evaluer leur fiabilite. Des essais en reacteur ont montre que les flux de neutrons rapides aggravent le fluage des metaux. Les renseignements ainsi obtenus permettent a l'ingenieur-concepteur de prevoir les deformations dues au fluage. On decrit dans le present rapport, l'experience ^cquise, au Canada, avec ies tubes de force. Le programme d'irradiation de ces tubes a joue un role de premier plan dans le developpement sans cesse grandissant du programme electronucleaire canadien.
Communication presentee au Troisieme Congres interamericain sur la technologie des materiau:-: (Rio de Janeiro, Bresil, 14—17 aout 1972).
L'Energie Atomiquedu Canada, Limitee Laboratories Nucleaires de Chalk River Chalk River, Ontario aout, 1972
AECL-4262 INTRODUCTION tube and is joined to the steel end-fittings. A cut-away of the Pickering power reactor, which has Pressure tubes are the pressure vessels of Canadian 390 fuel channels, is shown in Figure 2. designed CANDU (Canada Deuterium Umnium) nuclear reactors. They are about 10 cm in diameter, 0.3 to 0.5 cm thick by 6 m long and contain the fuel and cooling water operating at about 9.6 MN/m2 (1400 psi) and 300°C. A power reactor basically consists of many identical pressure tube assemblies. The Nuclear Power Demonstration reactor (NPD), the first CANDU reactor, was originally designed with a steel pressure vessel. In 1956, however, the design was changed to that of a pressure tube reactor, so that future CANDU reactors would not be restricted in size by the technological limitations of large steel pressure vessels. Since then we have come to appreciate the many other benefits, in design, manufacture, and safety, associated with the relatively small and very simple pressure vessel. Of major importance has been Canada's ability to gain experience by the irradiation of pressure tubes at full power reactor conditions in the test reactors [lj. Tube irradiations and evaluations have provided the information and confidence needed to proceed with an ever expanding power reactor program.
THE PRESSURE TUBE REACTOR INQ SHI11O COO I
\HD JMIIlO IIT 111 The Canadian reactors are designed for good :HO« fiA'i I StflllC "HO neutron economy; they use heavy water as moderator 3 Mil MAI Mnl
and neutron economic zirconium alloys in the core C*L*NDIiA mtLl II SHIlt imeiD CALAHD^IA &HIIL SHLILDS 3b Hflru** «
NPIl I pickKint and copper, depending on the extrusion temperature within the range 650-850°C. Copper improves the fc J PIIW HLW L'ooluni • HW" i extrudability of the billets and prevents excessive lli-d.l..r [...I,, r irsl ¥.<* W! ].'7J oxidation, while the steel prevents alloying of zir- <•"•-""• conium and copper at high extrusion temperatures. UN ,
v™,p. <• •2711 The extrusions are chemically declad and abrasively cleaned. Strength is developed either by cold crawing lr,,,m.l DiiimrK.r about 20% [6], or by quenching from 880°C and
Wall 1 h aging at 500°C [7]. During extrusion, the hexagonal zirconium develops strong crystallographic textures [8], which car. be controlled by the fabrication schedule. l>t".!|;ii Sin-i I'-1 PRESSURE TUBE PROPERTIES [.udirc huh The strength properties of zirconium allo- are anisotropic, i.e., they differ in the three principal directions according to the crystallographic texture, which depends on fabrication history. Typical tube properties are listed in Table II [9]. •rki-d /.r l!.:i wi ; Mi.
.•uUtl Zr-11.5 wt; Mb. TABLE II. PRESSURE TUBE PROPERTIES AT 300"C
fuelling, and th-^ channel is shop assembled. In both Matt-rial Testa 0.2% Yield Ultimate Burst Elongation Reductinn Strength Tensile Strength |%) in Area designs, the pressure tubes can be identical with only psi Strength psi the end-fittings different. (MN/.n;) (MN/mJ) (MN;m!)
The first material to be used for pressure tubes was LT •15,000 5-1,000 26 55 Zircaloy-2, an alloy of Zr—1.5 wt% Sn with small 13101 1372) TT 50,000 52,000 •u 54 additions of iron, nickel and chromium [5J. 13-151
Zircaloy-2 is also extensively used as fuel sheathing B 611.01)0 C3.000 28 35 and core structural components in many reactor 14111 (435) systems. However, Canada's interest in neutron Cold-Worked LT 53,000 76,000 15 50 /r 2 5 wt'. Nb |3«5| (524) economy led to the development of the stronger TT 77.000 81,000 23 54 Zr—2.5 wt% Nb alloy; tubes in the cold-worked [6J 1530) (559) and in the heat-treated [7j conditions have been used U 75.000 85.000 3-7 30 in CANDU power reactors. 1517) (585| MealTreated LT 69,000 86,000 19 61 Zr-2.5»-f; Nb (475| (5931
FABRICATION OF PRESSURE TUBES TT 95,000 98.000 12 50 (655| 1675)
The zirconium-alloy pressure tube is of simple B 92,000 109,000 2-5 20 shape, thin section, and homogeneous; the fabrica'ion (635) (7511 process is repetitive and subject to close control by " Ai)breviat:onr, -LI IonsiLudinal tensile;TV. transver*' tensile process parameters. Tubes are purchased to stringent . B. burst. specifications. Dimensional tolerances, freedom from defects, and surface finish can be guaranteed by 100% Development and pre-production tubes are irra- nondestructive testing of the finished tube. Chemical diated in the test reactors at power reactor composition, corrosion resistance and strength are conditions. Their behaviour is monitored periodically guaranteed by destructive tests on pieces cut from the and finally they are removed for destructive evalua- ends of the actual lubes used in the reactor. tions including burst tests [10,11,12]. Irradiation Ingots are made by multiple vacuum arc melting of generally increases the strength and reduces the electrodes of zirconium sponge compacted with the ductility. Typical properties of irradiated tubes are alloying elements. They are hot forged to billets given in Table III. The results from strength and other approximately 20 cm diameter and 45 cm long, tests are applied to the design of tubes for power which are either concentrically drilled or hot pierced reactors and to an assessment of their safety. -2- TABLE 111. BURST TEST PROPERTIES OF IRRADIATED PRESSURE TUBtS AT 300 c Tube irradiations have provided the information that the designer can apply to the relatively simple Material Exposurt- 0.2'fc Yiold Burst Elongation Riducliun pressure tube geometry and with a reasonable degree n/cnr St'imRlh Strength (%) in Arua of accuracy calculate the three-dimensioi.al creep P>1 MeV| psi psi i%) (MN/m-1 (MN/m-') deformations [ 13,14 j.
:I At present, creep due to internal pressure is of Culd-VYiirki'd 1.0 X 1U 7C.UUU 78,5uO 7 H \2 17 /ircalny-2 I5HH) 15111 most concern. When the pressure tube diameter
Culd-Wurked o.Kx in11 110.000 11! ,000 *1 1 5 - 17 increases by about 3%, coolant will tend to bypass Zr- 2 5 •»'!'. .Xb 17581 (7881 the fuel and decrease the heat transfer from fuel to
Heal-Tn'uU'd 1.2 X 10!l 144,000 157.000 ^1 14 coolant. Excessive deformation can be avoided by Zr-2.5 wt'.i 19921 (10801 designing for a low stress or by using more creep resistant material. The Zr—2.5 wt% Nb alloy has PRESSURE TUBE DESIGN BASIS about twice the creep resistance of Zircaloy-2 at PHW coolant conditions and thus represents a valuable Although zirconium alloys have not been included improvement. The effect of time on in-reactor creep in the ASME Boiler and Pressure Vessel Codes, the is not yet well established. The creep rate appears to pressure tubes for all CANDU reactors have been be decreasing with time; deformation calculations for designed to the intent of the Codes. One-third ce the the Zr—2.5 wt% Nb tubes in Pickering 3 and 4 and ultimatp tensile strength ior the unirradiated tube is Bruce reactors (designed for about 2'r diameter the design criterion even though the burst strength, p change in 30 years) assume no decrease in rate with or unirradiated or irradiated tubes, is higher. Design time. stresses are listed in Table I.
The stress distribution in the pressure tube can be CORROSION readily calculated since the tube is of very simple Experience from the irradiation of small specimens geometry. The tube is subject to internal pressure and pressure tubes in the test reactors has demon- and, for the horizontal tubes, to bending loads due to strated the efficacy of minimizing corrosion by the weight of fuel and coolant [13]. The ends of the controlling the coolant chemistry [9J. In pressurized pressure tube are rigidly joined to the end-fittings by water, radiolytic oxygen production is suppressed by rolled joints. The end-fittings are firmly supported by adding 5—10 cm3 H /kg D O. In a boiling coolant the end shield; the pressure tube is then a fixed beam 2 2 most of the hydrogen wouid pass into the gas phase, subject to a uniformly distributed load, and to point so ammonia is used to inhibit oxidation. Metal loss loads provided by the central spacers which in turn due to oxidation is predicted to be less than 5 x 1G~3 are supported by the calandria tube (see Figure 1). mm/year. Also slight wear occurs because of the Since there are no welds or appendages, solutions to sliding of the fuel along the pressure tube. The stress analyses are almost exact. combined corrosion and wear allowance is about 0.2 CREEP mm. Fast neutron flux enhances creep deformation, Deuterium and hydrogen are produced by the and it is the designers' responsibility to accommodate corrosion reaction in heavy water and light water, creep deformation and provide assurance of safe respectively. Some of the deuterium or hydrogen is operation. The tube irradiation program supple- absorbed by the zirconium alloy and, in excess of the mented by numerous small specimen creep tests has solid solubility, will precipitate as platelets of shown that at design conditions: Zr(D,H), 5. Early in the development of zirconium alloys we were concerned about hydrogen embrittle- — primary creep strain is relatively small and at a ment of pressure tubes. Later work showed that given temperature the & nndary creep rate is hydrides were not brittle above about 150° C [ 18j almost linearly proportional to stress and to the and that tube properties were not significantly fast neutron flux [9], affected by up to 400 ppm1 hydrogen (or 800 ppm — creep deformation is anisotropic with the best deuterium). Analysis of full-size pressure tubes that creep resistance in the hoop direction [14J, have been irradiated under power reactor conditions for up to 29,000 hours have shown that deuterium/ — high uniform elongation occurs. Deformations up hydrogen pickup will not impair the properties during to 9% have been observed in uniaxial, in-reactoi- the lifetime of the reactor [ 10,11,12j. creep tests [15,16] and, indeed, the deformation characteristics are such tnat "superplastic" behaviour is predicted [9,17], parts per million, by weight -3 IN-SERVICE INSPECTION valuable by supplying many tubes, highly irradiated under full service conditions, which could be artifi- The techniques for conducting inspections of cially defected and burst tested (see Figure 4). Only power reactor pressure tubes were developed in the when the pressure vessel is a tube are replicate burst test reactors, where development tubes are routinely tests on full-size components conceivable. The inspected [19]. Inspections are done when the artificial defects consisted of thin slits cut longitu- reactor is shut down and the fuel and coolant are dinally in the tube wall to represent cracks per- removed from a particular tube. Surface condition pendicular to the direction of maximum tensile stress. and any surface damage are determined by borescope Failure stress falls with increasing slit length (Figure and other instruments. Creep deformations are deter- 5); the slit length that leads to unstable crack mined from periodic measurements by diameter propagation at the pressure tube design stress is the gauges and straightness probes incorporating linear critical crack length for the alloy condition and test variable differential transformers. Other types of temperature. For current zirconium ailoys at 20 and nondestructive test equipment car. be used; an ultra- 300°C, the critical crack length is about 10 cm at the sonic instrument for wall thickness measurement was design stresses of CANDU reactors (110—180 developed but is seldom used. Eddy-current or MN/m2, 16000-26000 psi) and is unaffected by ultrasonic instruments for crack detection have not neutron irradiation [10,11,12]. Hydrogen up to been found necessary. ^=400 ppm aas no effect on critical crack length at Inspections were conducted in representative tubes 300°C, but reduces it at 20°C to a minimum of 4 cm in NPD in 1963, 1965, 1967, 1968 and 1971. In in hpjt-treated Zr—2.5 wt% Nb. The big difference 1971, the inspection showed the tubes to be in between tube wall thickness (0.5 cm) and critical excellent condition after 50,000 hours of power crack length (10 cm) gives a high degree of protection operation. An inspection of three tubes in Gentilly in from the risk of unexpected tube failure. A growing early 1972 showed the tube surface to be free of crack penetrates the wall of a pressure vessel when its surface damage. Douglas Point is to be inspected late length is approximately twice the vessel wall thick- in 1972. Routine inspection of the Pickering reactors ness [20]. The pressurized coolant will then leak will be performed by the utility's own operators, who through a crack, and be detected in the gas annulus have already made some of the initial measurements. between pressure and calandria tubes. Thus if the critical crack length is more than twice the wall Periodic inspection is an accepted procedure, that thickness, the pressure vesse! will leak but the crack should show any serious deviation from predicted will not propagate unstably. This is the 'ieak-before- creep, corrosion or surface wear behaviour of the break' criterion developed from fracture mechanics pressure tubes. Information from surveillance pro- theory [21]. grams on unstressed irradiated specimens cannot approach the value of information gained by direct Zirconium pressure-tube alloys in thin sections measurements. cannot develop high elastic constraints at the tip of a crack: extensive localized yielding redistributes the SAFETY applied stress. Brittle fracture is difficult to induce, Damage to a pressure vessel is always of concern. and a Kjc (minimum) fracture toughness value of 42 3 2 The tube inspection program has shown that damage MN.nf / (40,000 psi\/in.) was obtained in to a tube by fuel movement is negligible, but damage Zircaloy-2 only at sub-zero temperatures [22]. The by some of the special experimental assemblies or high values of fracture toughness give further confi- special tools has been found. Figure 3 shows one of dence in the safe behaviour of pressure tubes. these damaged areas. Such tubes are usually burst tested, and results show that shallow surface damage TUBE LIFETIME does not significantly decrease the tube strength [9]. The life of a pressure tube is at present considered However, further assurance was needed to demon- to be limited by creep deformation, which will lead strate the ability of tubes to tolerate severe damage. to increases in diameter from 2 to 3% in periods of Failure by overpressure of greater than three umes 20-40 years depending on material, thickness and design pressure is incredible because of the control operating conditions [9,13]. The limit is based on safety devices and overpressure relief systems in the possible deterioration in heat-transfer characteristics coolant circuit. Failure is credible only if a tube is due to coolant flow bypassing the fuel; the limit is damaged sufficiently to reduce its burst strength to not set by the material's ductility or tendency to the operating stress of the tube. rupture.
Again the tube irradiation program proved Since creep deformation can be reduced by -4- reducing the operating stress, the design life of the ment tubes of Zr-2.5 wt'c Nb were installed as an pressure tubes in a reactor is determined by the economical means of gaining experience with a new relative emphasis the designer places on neutron alloy in & power reactor. One pressure tube and one economy, station lifetime, and the costs and benefits calandria tube were replaced in the Douglas Point of retubing a complete reactor after many years of reactor because of damage to the calandria tutu- that service. had resulted from faulty installation of a booster rod. Retubing can be done. Many tubes have been Thus it has already been demonstrated that, iC installed, irradiated, and removed from the Canadian necessary, even the "pressure vessels" of CANDU test reactors NRU, NRX and WR1. Two Zircaloy-2 reactors can be replaced with relatively little tubes were removed from NPD in 1967; the replace- difficulty.
Figure 3 — Surface damage on the inner wall of a Zircaloy-2 pressure tube; (left) viewed from inside the tube through a borescope; (right) epoxy replica. White line is parallel to longitudinal axis of tube. -5- Figure 4 — (Left) Irradiated Zircaloy-2 pressure tube after burst test in a shielded cell: tube has bulged and split open. (Right) Assembling the burst test equipment in a shielded cell. coolants. There is interest in new alloys, e.g., zir- conium with additions of tin, molybdenum and -2 MN. m psi x 10" niobium, which have even higher strength than the binary Zr-2.5 wt% Nb alloys [23]. Those with low corrosion resistance can be clad with a corrosion- 300°C resistant zirconium alloy. Four clad pressure tubes have already operated in the test reactors, and their properties are being assessed for possible use in power reactors. 150 -22 The pressure tube concept has been proved to be sound and reliable: future development is primarily aimed at economic improvements 124 |. if) 100 -15 if) SUMMARY bJ Zirconium-alloy pressure tubes have been developed as the pressure vessels of CANDU reactors. cr Problems related to design, fabrication and inspection COLD WORKED Zr-2.5Wt%Nb^ en -—HEAT TREATED Zr-2.5Wt%Nb were solved with relative ease because of the simple -COLD WORKED ZiRCALOY-2 shape of the tube. cr Of most value has been the experience gained on 400 -58 pressure tubes operated at power reactor conditions 20°C in the test reactors. In-service inspection and post- irradiation destructive testing of representative pressure vessels has provided confidence in the safety aspects of the pressure tube concept. The irradiation program showed that creep was significantly enhanced by fast neutron flux, but it also provided the information the designer needs to accommodate creep deformation if the design life of a reactor is to be attained. With 2000 pressure vessels in service, there is ever-increasing experience and experimental data 100 -15 available to assist in the design of each new reactor.
2 3 4 5 in ACKNOWLEDGEMENT L I ,1.1 II The authors acknowledge the helpful comments of 2.5 5 7.5 10 12.5 cm J.A.L. Robertson.
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