Ignalina Source Book

3.2 Design of steel structures

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3.2.1 Standard Practices

 

Requirements for materials inside the core, where they are subject to intensive neutron fluxes, and for materials of equipment and pipelines, which are not exposed in such a way, are quite different. All metals, which are used for equipment and pipelines of the nuclear power plant, must have excellent mechanical properties, such as high corrosion and erosion resistance, certain thermophysical and good technological properties. It means, they must have an ability to withstand deformation without cracking under cold and hot conditions as well as good weldability and machinability.

For materials inside the core, the cladding and pressure vessels present additional requirements such as low cross neutron capture cross sections. This is needed for supporting chain reactions, as well as for improved mechanical properties, especially for strength against cleavage fracture and resistance to plastic deformation upon intensive neutron exposure.

Neutron exposure can result in damage of metal and alloy crystal lattice. Each metal structure transformation involves a change of its mechanical and thermophysical properties. Metals subjected to neutron exposure, become tougher and more brittle, their thermal conduction is decreased, and their creep rates are increased.

Thus, main requirements for construction materials are:

• corrosion and erosion resistance of coolants for specified parameters, compatibility with fuel and fission products,
• satisfactory mechanical properties (strength, plasticity, creep) taking into account radiation, which can change these properties,
• high thermal conductivity,
• low cross-section for neutron capture,
• technological effectiveness, such as machinability and weldability.

Note that, compared to the VVER-type reactors, the requirements for materials of metal structures of RBMK-type reactors is less restrictive. It is possible to use simpler low-alloy steel because of the lower coolant pressure, neutron fluency and thickness of different construction elements.

 

3.2.2 Material Properties

 

The RBMK-1500 reactor is located in a cylindrical casing made from a sheet of heat-resistant low-alloy steel 10ChN1M, the thickness of which is 16 mm. Reactor casing, together with the top and bottom metal structures, the thickness of which is 40 mm and is manufactured from the same material, form the close reactor space. Chemical composition, physical-mechanical and thermophysical properties of the low-alloy steel are given in Tables 3.2 through 3.4.

Maximum permissible temperature for steel 10ChN1M is 400 ^oC [17]. This steel is not prone to thermal fracture by aging in the temperature range of 340-450 ^oC and duration time up to 10⁴ hours. Steel 10ChN1M is sufficiently effective to all metallurgical conversions. The statistical data of mechanical tests of specimens show that the material is stable and has a high level of strength, plasticity and viscosity.

For welded structures, which are manufactured without a thermal treatment, the ability of steel to relieve stresses during operation is very important. At a temperature of 400 ^oC and the duration time of 10³ hours an initial stress equal to 0.8 x s_y is reduced about 10 % and at temperature of 450 ^oC - by about 30 % [23].

Steel 10ChN1M with a content of nickel up to 1.4 % ensures full hardenability of sheets and, hence, uniform mechanical properties for thickness not more than 50 mm. The critical fracture temperature for steel 10ChN1M sheets for thickness of 30-40 mm is about 10 ^oC [23]. This steel has a high resistance for thermal fracture, and maximum rise of critical fracture temperature depends somewhat on aging temperature and the critical fracture temperature 10-20 ^oC [23]. Steel 10ChN1M has good radioactive resistance to fluency up to 6× 10¹³ n/cm² (E > 1 MeV) [21].

Main circulation circuit of the RBMK-1500 reactor is manufactured mainly from austenitic stainless steel 08Ch18N10T. Pressure and suction headers with corresponding pipelines, as well as the separator drum, are manufactured mainly from carbonic steel 22K, and the lining inside by austenitic stainless steel 08Ch18N10T. Chemical composition, physical-mechanical and thermo-physical properties of mentioned steels are given in Tables 3.2 through 3.4.

 

Table 3.2 Chemical composition of steels, used for the main equipment of RBMK-1500 reactors [23]

 

Steel type

Chemical composition, %

C

Si

Mn

Cr

Ni

Ti

Mo

S

P

Cu

10ChN1M

0.08-0.12

0.17-0.37

0.3-0.6

0.7-1

1.1-1.4

-

0.4-0.6

< 0.03

< 0.03

< 0.3

08Ch18N10T

< 0.08

< 0.8

< 2

17-19

9-11

< 0.7

-

< 0.02

< 0.035

-

22K

0.19-0.26

0.2-0.4

0.75-1

< 0.4

< 0.3

-

-

< 0.025

< 0.025

< 0.3

 

Table 3.3 Physical-mechanical properties of steels, used for the main equipment of RBMK-1500 reactors [21]

 

Steel type	Characteristic Specification		Temperature, oC
			20	50	100	150	200	250	300	350	400	450	500	550
10ChN1M	Sheet	s u, MPa	540	530	520	500	491	491	461	451	441
	thickness	s y, MPa	441	432	422	412	402	392	373	363	353
	from 6 to 40 mm	e u, %	16	16	16	13	13	13	13	13	14
		A, %	50	50	50	50	45	45	40	40	40
	Weld-less tubing	s u, MPa	491	491	471	451	441	432	412	402	392
	hot rolled with	s y, MPa	343	333	323	314	304	294	294	294	275
	outside diameter	e u, %	20	20	20	15	15	15	15	15	15
	60-168 mm and wall thickness from 6 to 32 mm	A, %	50	50	50	50	45	45	40	40	40
08Ch18N10T	Rolled steel	s u, MPa	491	480	461	436	417	397	377	353	328	314	289	270
	and forging	s y, MPa	196	191	189	186	181	176	172	167	162	157	152	150
	with thickness or	e u, %	38	37	36	33	31	28	26	25	22	20	20	20
	diameter less than 200 mm	A, %	40	40	40	40	40	40	40	40	40	40	40	40
	Plates, forging	s u, MPa	491	477	456	426	417	382	358	333	303	289	260	235
	from ingot, sheet	s y, MPa	196	193	186	181	176	167	162	157	152	144	137	132
	bar and stamping	eu, %	35	34	33	31	29	27	26	25	24	23	22	22
	with diameter from 40 to 200mm	A, %	40	40	40	40	40	40	40	40	40	40	40	40
	The same,	s u, MPa	491	475	446	421	392	368	343	314	289	260	235	206
	with diameter	s y, MPa	196	191	181	172	164	152	147	137	132	123	113	103
	more than	e u, %	35	34	33	31	29	27	26	25	24	23	22	22
	200 mm	A, %	40	40	40	40	40	40	40	40	40	40	40	40
	Piping	s u, MPa	510	471	461	441	421	421	412	412	402	382	353	333
		s y, MPa	216	206	206	196	187	187	177	177	167	157	147	147
		e u, %	35	32	30	28	27	26	26	26	25	25	25	25
		A, %	55	55	55	54	54	53	52	51	50	48	47	45
22K	Sheets with	s u, MPa	430	430	430	430	430	421	412	392
	thickness from	s y, MPa	215	206	196	186	186	186	186	177
	70 to 170 mm	e u, %	18	18	18	17	17	16	17	18
		A, %	40	40	39	38	38	38	39	40
	Forging with	s u, MPa	390	390	390	383	373	363	353	353
	diameter from	s y, MPa	195	186	177	167	167	157	157	137
	300 to 800 mm	e u, %	18	15	13	13	13	13	13	13
		A, %	38	38	38	36	36	35	34	34
	Forging with	s u, MPa	430	392	392	392	392	392	353	343
	diameter from	s y, MPa	215	206	196	186	186	186	186	177
	100 to 800 mm	e u, %	16	14	11	11	11	11	11	11
		A, %	35	35	35	33	33	32	31	31

s _u - ultimate strength, MPa,

s _y - yield strength, MPa,

e _u - ultimate strain, %,

A - area reduction, %.

 

Table 3.4 Thermo-physical properties of low-alloy steels [23,24]

 

Temperature, oC	10ChN1M			08Ch18N10T				22K
	a , 10-6/K	E, GPa	a , 10-6/K		E, GPa	l , W/(mK)	a, 10-6/K		E, GPa	l , W/(mK)
20	-	210	-		205	-	-		200	-
50	11.5	207	16.4		202	-	11.5		197	-
100	11.9	205	16.6		200	16.3	11.9		195	49.5
150	12.2	202	16.8		195	-	12.2		192	-
200	12.5	200	17.0		190	17.5	12.5		190	47.7
250	12.8	197	17.2		185	-	12.8		185	-
300	13.1	195	17.4		180	18.8	13.1		180	45.5
350	13.4	190	17.6		175	-	13.4		175	-
400	13.6	185	17.5		170	21.4	13.6		170	43.5
450	13.8	180	18.0		167	-	13.8		165	-
500	14.0	175	18.2		165	23.0	14.0		160	41.5
550	14.2	170	18.4		162	-	14.2		-	-
600	14.4	165	18.5		160	24.6	14.4		-	39.3
700	-	-	-		-	26.8	-		-	-

 

a - coefficient of thermal expansion, 1/K,

l - thermal conductivity, W/(mK),

E - Young's modules, GPa.

 

Table 3.5 Steel 08Ch18N10T properties at 20 ^oC after exposure to fast neutrons at different fluences [24]

 

Fluency, n/cm2	s u, MPa	s y, MPa	e u, %
At exposure temperature of 100 oC
-	675	340	53
10E17	730	380	48
5x10E17	720	530	45.5
4.3x10E18	875	710	37
9x10E18	780	680	34.5
10E20	880	780	23
At exposure temperature of 300 oC
-	550	-	40
3x10E19	650	-	27
6x10E19	680	-	21
10E20	720	-	18
3x10E20	750	-	14
5x10E20	770	-	18
7x10E20	790	-	12

 

Maximum permissible temperatures for steels 08Ch18N10T and 22K are 600 ^oC and 350 ^oC, respectively [22]. Austenitic stainless steel 08Ch18N10T possesses high heat resistance right up to 550-600 ^oC [24]. This steel has high uniform corrosion resistance in the water of main circulation circuit up to 360 ^oC, and in water-steam mixture - up to 650 ^oC. Losses due to corrosion at temperatures 260-315 ^oC amount to 0.0003-0.0018 mm/year [25].

Cross-section of heat neutron absorption for steel 08Ch18N10t is (2.7-2.9)× 10^-28 m². Neutron fluency influence to physical-mechanical properties of austenitic stainless steel is shown in Table 3.5.

With an increase of neutron fluency the elongation of steel 08Ch18N10T decreases. By exposure to temperatures higher than 350 ^oC partial annealing of the radioactive effects and, hence, reestablishment of physical-mechanical properties of steel 08Ch18N10T take place. The temperature of the annealing is about 0.5-0.55 of the absolute melting temperature. Exposure of austenitic stainless steel to temperatures higher than 600 ^oC leads to loss of fragility.

Neutronic exposure has practically no influence on the corrosion of steel 08Ch18N10T. It can be subject to corrosion cracking with simultaneous presence of chlorine and oxygen or other oxidizers and with the presence of stretching of the metal. If the stressed metal is in the surroundings, of only 1 % of the mentioned agent, cracking is not observed [24]. Corrosion cracking of the steel 08Ch18N10T is observed with the concentration of chlorides on the surface. It is observed that, if concentration of the chlor-ion on the surface of steel is in saturated steam which contain 0.3-0.4 g/kg oxygen, corrosion is 10^-6 g/kg.

Steel 22K is sensitive to thermal fracture of the operating temperature of the NPP equipment. Maximum rise of the critical temperature of fracture depends very little on the temperature of aging. For heating under extended time of 10⁴ hours it make up 10 ^oC at aging temperature of 340 ^oC and 20 ^oC at aging temperature of 400 ^oC [23].

As a construction material for manufacturing of cladding of fuel assemblies and fuel channels of the RBMK-type reactors zirconium alloys with admixture of niobium are used. Zirconium and its alloys have a small cross-section of absorption of heat neutrons equal to (0.2-0.3) × 10^-29 m². Zirconium keeps satisfactory mechanical properties at temperatures up to 390-400 ^oC. Zirconium practically does not interact with nuclear fuel, does not undergo plastic deformation either under hot or cold conditions, and has good weldability. Zirconium and its alloys have good corrosion resistance to reactor water at temperatures up to 370 ^oC.

If water possesses both oxygen and ammonia, corrosion resistance visibly decreases. This condition imposes definite limitations on the choice of the water regime.

At temperatures of 1200-1300 ^oC zirconium actively interacts with water. This process leads to the release of a large amount of heat. This condition is taken into account, when analysis of accident situations are performed.

Niobium not only improves mechanical properties of the alloy, but also neutralizes bad influences of admixtures to the corrosion resistance.

Rise of hydrogen content in zirconium and its alloys evoke a hydrogen-induced fracture, which is displayed mainly by a decrease of impact strength - at 20 ^oC it can decrease by 4-6 times. Hydrides have little influence to indexes of statistical strength and strain elongation.

Cladding of fuel assemblies use iodide-zirconium alloy with 1 % of niobium, and cladding of fuel channels use an alloy with 2.5 % of niobium. The physical-mechanical properties of the mentioned alloys as well as the Zyrcaloy-2 properties, are shown in Table 3.6. Maximum permissible temperature for using zirconium alloys with 1 and 2.5 % of niobium is 360 ^oC [21].

Influence of neutron exposure on zirconium and its alloys is similar to the influence on steel: ultimate resistance and especially yield strength increases, ductility decreases, creep accelerates. Influence of neutronic exposure to fracture of zirconium alloys with niobium is lower in comparison with that of Zyrcaloy-2. With rise of temperature up to 350-400 ^oC bad influence of exposure to creep decreases. Influence of exposure to physical-mechanical properties of alloy Zr+2.5 % Nb is shown in the Table 3.7.

Corrosion process are accelerated with increasing amounts of oxygen in the reactor water and the availability of exposure. Without exposure such a process was not observed. The influence of exposure to hydrogenation is less than that to oxidation. Hydrogen absorption by zirconium alloys with niobium is less visible than that by Zyrcaloy-2.

In the wall of the fuel channel of the zirconium alloy an accumulation of plastic deformation is observed during its life-time. Zirconium alloy with 2.5 % of niobium has good ductility at initial conditions. However, due to aging, exposure, creep and cyclic change of temperature, the ductility properties gets worse.

Fracture of fuel channels because of loss of long-term strength occurs at residual strain of 5-10 %. According to the studies of the exposure limit on long term strength increase [25]. Corrosion rate of alloy Zr+2.5 % Nb does not exceed 0.01 g/(m²hour) during the 8000 hour testing, and under reactor exposure conditions, rise by 5-10 %.

Aluminum alloys are used for the CPS equipment. They have good corrosion resistance in the water at temperatures 100-250 ^oC. Maximum permissible temperature for these alloys is 190 ^oC [21]. Ultimate strength of these alloys at 20 ^oC is 150 MPa, and the yield strength is 40-50 MPa. At a temperature of 200 ^oC these indexes decrease to 90-100 and 30-40 MPa, respectively.

 

3.2.3 Failure Design Criteria

 

Similar to the design of other reactor types, the design of RBMK-1500 reactors employs a ‘safety factor’ in the determination of the strength of equipment and pipelines. Maximum stresses in structures must not exceed the permissible stresses. Permissible stresses are determined from the material characteristics at recommended temperatures and include the safety factor. In the code for strength calculations [21] the following recommended temperatures were given:

• 20 ^oC for aluminum and titanium alloys,
• 250 ^oC for zirconium alloys,
• 350 ^oC for carbon, alloy, silicon-manganese and high-chromium steels,
• 450 ^oC for corrosion-resistant austenitic and high-temperature chrome-molybdenum-vanadium steels and for ferrous-nickel alloys.

 

Table 3.6 Physical-mechanical zirconium alloy properties [24]

 

Alloy	Characteristics	Temperature, oC
		20	200	300	400
Zr+1 % Nb	s u, MPa	350	260	200	180
	s y, MPa	200	160	120	90
	d , %	30	31	33	38
Zr+2.5 % Nb	s u, MPa	450	320	300	270
	s y, MPa	280	220	200	180
	d , %	25	24	23	22
Zyrcaloy-2	s u, MPa	480	250	200	170
	s y, MPa	310	150	100	70
	d , %	22	34	35	36

d - plasticity, %.

 

 

Table 3.7 Influence of fluency exposure on physical-mechanical properties of alloy Zr+2.5 % Nb [24]

 

Working	Fluency, n/cm2	Test temperature, oC	s u, MPa	s y, MPa	e u, %	d , %
Hardening	-	20	870	780	63	13
with 880 oC	1020	20	1000	960	-	10
and aging	1021	20	1110	1080	45	8
with 500 oC	-	300	580	530	75	14
during	1020	300	720	680	-	13
24 hours	1021	300	810	780	65	9
Hardening	-	300	580	480	70	13
with 960 oC	1020	300	810	770	50	8
and aging	1021	300	860	860	5	4
with 500 oC during 24 hours

 

According to the same design code [21], nominal permissible stresses for equipment and pipeline components, which are pressure loaded, take the minimum of the following values:

s = min {s _u(T)/n_u, s _y(T)/n_y, s _ut(T)/n_ut},

where s _u - ultimate strength, s _y - yield strength, s _ut - minimum protracted strength during time period t.

For components of equipment and pipelines, which are loaded by internal pressure,

n_u = 2.6, n_y = 1.5, n_ut = 1.5.

For components of equipment and pipelines, which are loaded by external pressure, which is higher than the internal pressure

n_u = 2.6, n_y = 2, n_ut = 2.

This calculation procedure for permissible stresses is used for static loading conditions. Structural components subjected to variable, cyclic loading, as a rule, are damaged by lower stresses. Therefore, for the final design strength and stability conditions, final calculations are performed, which take into account all the expected loading and all the operation regimes. During these final calculations, the following characteristics are taken into account:

• static strength,
• stability,
• cyclic and long-term cyclic strength,
• resistance to cleavage fracture,
• prolonged statistical strength,
• progressing deformation,
• influence of seismic loading,
• resistance to vibration.

The final calculation procedure is described in detail in [21], pages 45-119.

 

3.2.4 Qualification Tests of Reactor Components

 

The materials of the equipment, including the materials used in the reactor core, must maintain during its operational life-time high strength with a sufficient level of ductility and high corrosion resistance. In order to avoid destructive failure the NPP, equipment is tested before installation for expected conditions during its life-time. The design, manufacturing, mounting and operation of the main metal components and their welds would be subjected to the requirements, which are regulated by the references [23, 27, 28]. These requirements are extended to vessels under pressure (including hydrostatic and vacuum), reactor vessels with their guard tanks and casings, pump vessels, pipelines and devices of first and second circuit of the NPP.

 

Hydraulic Tests [23]

 

Hydraulic tests are performed to control the strength and leakage of equipment, which is loaded by pressure. Hydraulic tests of this equipment are performed after the manufacturing and installation has been completed.

Pressure of the internal hydraulic tests (P_h) must not be lower than those calculated by the formula:

P_h = K_h P s (T_h)/s (T) (lower bound),

and not higher than the pressure which would cause the tested equipment a nominal membrane stress to 1.35 s (T_h), and the sum of the nominal local membrane and the nominal flexural stresses to 1.7 s (T_h) (upper bound),

where, K = 1.25 for equipment and pipelines, and K = 1 for confinement and for guard tanks,

P is allowable pressure during tests at the manufacture plant or working pressure during tests after installation and operation, MPa,

s (T_h) is the nominal allowable stress at temperature of the hydraulic tests T_h, MPa,

s (T) is the nominal allowable stress at recommended temperature T, MPa.

For components, which are subjected to loading by external pressure, the following condition must be fulfilled P_h < 1.25P.

For a pressure P up to 0.49 MPa, the value of P_h should be more than 1.5P, but not less than 0.2 MPa. For a pressure P above 0.49 MPa, the value of P_h should not be less than (P+0.29) MPa.

Hydraulic tests of equipment would be performed at the temperature of the tested medium, where the metal temperature of the tested equipment is not lower than the minimum allowable, as calculated by the code [21]. In all cases the temperature of the test and the surrounding medium can not be lower than 5 ^oC. It is also possible to calculate the minimum permissible metal temperature T_h during a hydraulic tests from the correlation’s:

T_h > T_k0 - 260 + 73× 10^-6Ss ²_y, if Ss ²_y< 3.5× 10⁶,

T_h > T_k0 - 17 + 3.1× 10^-6Ss ²_y, if 3.5× 10⁶ < Ss ²_y < 25× 10⁶,

T_h < T_k0 + 48 + 0.47× 10^-6Ss ²_y, if Ss ²_y > 25× 10⁶,

where, T is the critical temperature of metal fracture at initial conditions, ^oC,

S is the maximum nominal wall thickness of the equipment, mm,

s _y is a limit of material yield stress at temperature 20^oC, MPa.

Pressures as well as temperatures of the hydraulic tests conducted after manufacturing, are indicated by the manufacturer in the equipment certificate.

The permissible metal temperature in the hydraulic tests during equipment operation is determined by the owner of the equipment on the basis of calculated strength, data from the equipment certificate, number of load cycles (which is known from the operational process) neutron fluency with energy E > 0.5 MeV and the surveillance-specimen test data.

The exposure time for the equipment subjected to pressure P_h during hydraulic tests must not be less than 10 minutes. After the exposure, the pressure is decreased to the value of 0.8× P_h and a visual inspection is performed wherever possible.

The measurement of pressure during the hydraulics tests has to be performed by at least two independent measuring channels. A measurement error can not exceed 5 % of the nominal value of the test pressure. Temperature must be controlled by devices with the sum error not higher than 3 % from the maximum value of the measuring temperature.

A test program is prepared before the hydraulic tests. After the test the data are prepared and documented.

The equipment is considered having passed the test, if during the test or visible inspection it does not reveal leaks or rupture of metal, pressure drop does not exceed a permissible limit, and if after the tests visible residual strains are not discovered.

 

Control of the State of Metallic Equipment During Operation [23]

 

Control of the state of the NPP metallic equipment is performed to uncover light structural defects, changes of physical-mechanical properties, as well as to estimate the metal. Estimation of the metal state during the equipment operation is performed using both indestructible and destructible methods. Metal testing by indestructible methods is performed using visual, capillary, supersonic and radiography methods. Destructible methods are used to test metal components and welds by means surveillance-specimens, namely:

• changes of mechanical properties (yield strength, temporary resistance, change in elongation, change in cross-section),
• characteristics of resistance of cleavage fracture (critical temperature of fracture, viscosity of destruction or critical crack opening),
• characteristics of solid and local corrosion (inter-granular stress corrosion),
• changes in characteristics of cyclic strength (fatigue strength).

Periodical inspection of metallic components by indestructible methods is performed on the following time scale:

a) first - not later than after 20000 hours of operation,

b) subsequent - not later than after 30000 hours from the previous test.

Tests of the surveillance-specimens, which are placed inside the reactor vessel, are performed not less than 6 times per expected life-time of the component. First unloading and tests of surveillance-specimen are performed after one year from the start of the operation, and subsequently - every three years during the first 10 years of operation on the condition, that during the first unloading the neutron fluency at the reactor vessel is not less than 10²² n/m², but not more than 10²³ n/m² (E > 0.5 MeV). Surveillance-specimens for control of mechanical properties and its characteristics for redundancy to cleavage fracture are placed into the fuel channels. Detectors for measuring of neutron fluency and temperature are placed with the specimens. Surveillance-specimens are made by the manufacturer of the equipment.

 

Technological Control of RBMK-1500 Reactor Components

 

In order to avoid reactor component damage it is necessary to ensure an optimal thermal operation regime. Technological control of reactor components is provided by a system of temperature monitors of graphite cladding and reactor metal structures.

Temperature of cylindrical casing of the reactor is controlled along one generatrix at four elevation points. On top and bottom metal structures, which are under considerable thermal stresses, 30 points are monitored during stationary and transient regimes. Cable chromel-alumel thermocouple (type TChA-1449) placed inside protective case of corrosion- resistant steel with two modification - for putting into another special case and one without that case - is used for temperature measuring of metal structures. The use of special cases for positioning the TChA-1449 thermocouple does make it possible to change thermocouples.

The temperature of metal structures is automatically monitored by the information-computing system "TITAN" and in the case of excess of an assigned value, a deviation is depicted on the screen and in print. Information about temperatures of the metal structure is available upon operator's request.

The error of the temperature measurement does not exceed 2 %, the time constant of the cable thermocouple is less than 5 s, and the time constant of the thermocouples in the protective case - no more than 60 s. The magnitude of cumulative errors are estimated for internal heat-evolution in elements of thermometric devices and for heat exchange conditions using benchmarked objects. As a rule, methodical errors have a positive sign. This means that thermal converters give an excessive reading in comparison with the actual temperature value. Thus, there is some reserve in the control parameters.

Primary converters of type D-2373 are used for control of pressure differences between the cylindrical casing and the reactor space.

 

Quality Assurance of Welded Joints

 

A large number of welds are present in an RBMK type NPP. Welding is the main method for joining a considerable number of pipelines with elements of equipment, devices and metal structures, which were manufactured from carbon, austenitic and other special steals. Pipelines at Ignalina NPP are of different outside diameters ranging from less than 0.1 to more than 1.0 m. Total amount of welded joints reaches hundreds of thousands. Therefore, a continuous control of quality assurance of welds by appropriate measuring equipment is necessary.

For quality control of welded joints and surfacing the following methods are used [28]:

• visual inspection and measurement of main dimensions of welds,
• chromatic and luminescent detection (to reveal internal defects, which become apparent on the surface),
• magnetic-particle inspection (for ferro-magnetic steals, it is possible to find defects at depths up to 8-10 mm from the surface),
• radiographic inspection,
• ultrasonic detection,
• hydraulic tests of strength and leakage,
• control of welded joint tightness by means of helium and halogen leak detectors,
• laboratory testing methods (mechanical properties, metallography investigations, tests of inter-granular corrosion, etc.). This inspection should be carried out for the surveillance-specimens that are cut-out from the welded joints made according to the same welding technology using the same surfacing materials and heating facilities, welded using the same equipment and subjected to the same monitoring method as the joint under inspection.

The specimens are placed in a special container. A complete set should include the following specimens:

• specimens for determining the mechanical properties,
• specimens for determining the critical fracture temperature.

Six containers each of which is a single set, represent a suspension that is arranged in the reactor close to the bottom and top metal structures of the reactor. The suspension structure enables the removal of specimens without difficulties and to separate each section for handling and further testing of the specimens.