System and method for in situ testing of the leak-tightness of a tubular member

System and method for in situ testing of the leak-tightness of a tubular member, which tubular member may be a nuclear steam generator heat transfer tube. The system includes a mandrel insertable into the tube, the mandrel having a pair of spaced-apart expandable bladders surrounding the mandrel. A first channel extends through the mandrel and is in communication with the bladders. A fluid supply circuit is in fluid communication with the first channel for supplying pressurized liquid to the bladders so that the bladders radially expand into sealing engagement with the inner diameter of the tube. As the bladders expand into sealing engagement with the tube, a sealed annular chamber is defined therebetween bounded by the bladders and the inner diameter of the tube. A second channel also extends through the mandrel, the second channel being in fluid communication with the fluid supply circuit at one end thereof and in communication with the chamber at the other and thereof in order to pressurize the chamber with the liquid. A controller is connected to the fluid supply circuit for measuring the flow rate of the liquid flowing through the fluid supply circuit to the chamber. Any crack in the tube wall adjacent the chamber will allow the liquid to escape therefrom at a characteristic flow rate corresponding to the size of the crack. However, the fluid supply circuit maintains the liquid pressure in the chamber so that the pressure and thus the liquid inventory in the chamber remains constant as the liquid leaks through the crack. The controller, which is connected to the flow meter, measures the flow rate and pressure of the liquid flowing through the fluid supply circuit. In view of the fact that the liquid (i.e., water) is incompressible, the flow rate through the crack and the pressure in the chamber are provided by measuring the flow rate and pressure of the liquid flowing through the fluid supply circuit.

BACKGROUND 
This invention generally relates to leak testing and more particularly 
relates to a system and method for in situ testing of the leak-tightness 
and leak rate of a tubular member, which tubular member may be a heat 
transfer tube of the kind found in typical nuclear steam generators. 
Although leak testing devices and methods are known, it has been observed 
that these devices and methods have a number of operational problems 
associated with them which make such devices and methods less than 
completely satisfactory for leak testing heat transfer tubes of the kind 
found in typical nuclear steam generators. However, before these problems 
can be appreciated, some background is desirable as to the structure and 
operation of a typical nuclear steam generator. 
In this regard, a typical nuclear steam generator or heat exchanger 
generates steam when heat is transferred from a heated and radioactive 
primary fluid to a non-radioactive secondary fluid of lower temperature. 
The primary fluid flows through a plurality of U-shaped tubes disposed in 
the steam generator. The secondary fluid flows across the exterior 
surfaces of the tubes as the primary fluid flows through the tubes. The 
walls of the tubes function as heat conductors for transferring the heat 
from the heated primary fluid flowing through the tubes to the secondary 
fluid of lower temperature flowing across the exterior surfaces of the 
tubes. As the heat is transferred from the primary fluid to the secondary 
fluid, a portion of the secondary fluid vaporizes to steam for generating 
electricity in a manner well understood in the art. 
Occasionally, due to tube wall cracking caused by stress and corrosion 
during operation (i.e., known in the art as primary water stress corrosion 
cracking), the steam generator tubes may degrade (i.e., experience tube 
wall thinning) and thus may not remain leak-tight. If through-wall 
cracking occurs due to the degradation, the radioactive primary fluid may 
undesirably leak through the crack and commingle with the nonradioactive 
secondary fluid. 
Therefore, such leaking tubes may be plugged to prevent commingling the 
radioactive primary fluid with the nonradioactive secondary fluid. 
However, for safety reasons, technical specifications imposed by the 
United States Nuclear Regulatory Commission on each nuclear reactor power 
plant holding an operating license set a limit on the percentage of heat 
transfer tubes that may be plugged. Exceeding this technical specification 
limit jeopardizes the operating license of the power plant. It is 
therefore prudent to plug only those heat transfer tubes that require 
plugging. 
However, it has been observed that not all leaking tubes require plugging 
because a small amount of tube leakage is tolerable. This is so because 
small amounts of tube leakage do not pose safety concerns. In view of 
this, the United States Nuclear Regulatory Commission has issued Draft 
Regulatory Guide 1.121 (NUREG-1477) which provides a basis for determining 
acceptable tube leakage in order to identify which degraded or leaking 
steam generator tubes must be plugged in order to satisfy safety 
requirements. In general, Draft Regulatory Guide 1.121 sets a limit for an 
acceptable leak rate at a specified pressure in the tube, the pressure 
being that which would be experienced during a postulated accident (e.g., 
main steam line break). Therefore, under the requirements of Draft 
Regulatory Guide 1.121, a degraded or leaking tube maybe "leak-rate 
tested" to determine whether or not the tube will have an acceptable leak 
rate during such a postulated accident. If the leak-rate is acceptable, 
then the leaking tube need not be plugged. Of course, it is desirable to 
perform such a leak-rate test in a precise and cost efficient manner. 
Techniques for determining the leak rate from degraded nuclear steam 
generator tubes are known. In this regard, it is known that as an eddy 
current inspection probe is translated along the inner diameter of a tube, 
its voltage amplitude will increase when it passes a degraded portion of 
the tube. The degraded portion of the tube may have a through-wall crack 
which will allow fluid to leak therethrough at a flow rate related to the 
size of the crack and the fluid pressure within the tube. That is, the 
voltage amplitude will have a statistical correspondence with the flow 
rate through the crack at a given pressure. Thus, by knowing the voltage 
amplitude and internal tube pressure, one can determine the flow rate 
through the crack. In other words, the voltage amplitude of the eddy 
current probe can be correlated to a given flow rate and pressure. It is 
therefore possible to make a plurality of empirical or experimental 
measurements of voltage, leak-rate and pressure to establish a data base 
of flow rate as a function of eddy current voltage and anticipated 
pressures. Such a database can then be used in the field to conveniently 
determine the anticipated leak rate at the pressure of interest merely by 
measuring the voltage amplitude, as explained more fully hereinbelow. 
Typical prior art methods of gathering the empirical data to establish the 
data base are discussed hereinbelow. 
It is important to precisely and cost-effectively determine the empirical 
data for the data base. However, two prior art methods typically used to 
obtain the empirical data for the previously mentioned data base are 
costly and may be imprecise. One of these prior art methods entails 
constructing a bench-scale model steam generator and then inducing 
through-wall degradation in the tubes that are disposed in the model steam 
generator. Eddy current voltage amplitude measurements of the degradation 
are then made by passing an eddy current probe through the tube. Next, the 
degraded tube is pressurized at a plurality of pressure values and the 
pressures and leak rates measured. The eddy current voltage amplitude, 
pressures and leak rates are recorded in a data base. However, a problem 
with this method is that the bench-scale model steam generator is 
constructed at a not insignificant expense and the empirical data obtained 
may or may not be truly representative of the real conditions in a 
full-sized steam generator belonging to an actual operating nuclear power 
plant. 
Another prior art method of establishing the data base entails inserting an 
eddy current probe into a degraded tube of a full-sized steam generator 
belonging to an actual power plant to obtain the required voltage reading 
and then removing the tube from the steam generator. Once the tube is 
removed, the tube is leak-rate tested in order to establish the previously 
mentioned leak rate and pressure values for the data base. However, 
applicants have observed that the process of removing the tube to be 
examined may result in a change in morphology of the degraded portion of 
the tube such that the morphology of the degraded portion of the tube 
after removal is not the same as before the tube is removed. This may 
occur, for example, when a network of cracks in the degraded portion of 
the tube are connected by ligaments which become torn during the removal 
process. Thus, the morphology of the tube after removal may be different 
than before removal. Applicants have confirmed this phenomenon by 
observing that eddy current voltage amplitudes are sometimes larger after 
removal than before removal. Therefore, this second prior art method of 
establishing the data base may cause the data base to be imprecise and is 
therefore not preferred. In addition, this second prior art method of 
establishing the data base cannot be conveniently performed in situ; that 
is, the tube must be removed. Moreover, the process of removing the tube 
is time consuming and necessitates that the steam generator betaken 
off-line for an extended period of time. Each day that the steam generator 
is off-line requires the reactor owner to incur approximately $1,000,000 
in replacement power costs. 
Therefore, a problem in the art is to provide a technique for steam 
generator tube in situ leak-rate testing that is precise and cost 
effective in order to establish the data base while overcoming the 
disadvantages of the prior art. 
After the data base is established, an eddy current measurement is 
performed in the field on a preselected steam generator tube to obtain the 
voltage amplitude measurement. Next, the voltage amplitude is looked-up or 
found in the data base to obtain an anticipated leak flow rate at an 
anticipated pressure (e.g., pressure during a postulated accident). In 
this manner, the flow rate through a leaking steam generator tube of an 
operating power plant during a postulated accident may be determined 
simply by performing an eddy current inspection of the tube and then 
correlating that eddy current voltage amplitude to flow rate and pressure 
data existing in the data base. However, in order to establish the values 
for the data base, a plurality of steam generator tubes first must be 
leak-rate tested. 
Yet another apparatus and method for testing the tightness of closed-end 
tubes in heat exchangers of nuclear reactors is disclosed in U.S. Pat. No. 
3,919,880 titled "Method And Apparatus For Testing Closed-End Tubes In 
Heat Exchangers of Nuclear Reactors And The Like" issued Nov. 18, 1975 in 
the name of Gunter Seyd, et al. This patent discloses a method and 
apparatus for detecting a leak in a heat exchanger tube of a nuclear 
reactor by inserting an expandable plug which seals off a closed region in 
the heat exchanger tube. A pressurized fluid is then introduced into the 
closed region, and the pressure of the fluid is monitored for detecting a 
leak through the closed end of the tube. However, this patent merely 
discloses testing for the leak-tightness of a closed-end tube with a fluid 
under pressure, and merely discloses sealing the tube at one point with an 
expandable plug and testing the area located between the expandable plug 
and the closed end of the tube to determine if the closed end of the tube 
permits leaking. This patent does not appear to disclose leak testing for 
fluid flow rate through a crack in the walls of an open-ended tube. Leak 
testing of an open-ended tube is required for leak testing steam generator 
heat transfer tubes in the manner satisfying Draft Regulatory Guide 1.121 
(NUREG-1477). 
Therefore, what is needed are a suitable system and method for in situ 
testing of the leak-tightness of a tubular member, which tubular member 
may be a heat transfer tube of the kind found in typical nuclear steam 
generators. 
SUMMARY OF THE INVENTION 
Disclosed herein are a system and method for in situ testing of the 
leak-tightness of a tubular member, which tubular member may be a nuclear 
steam generator heat transfer tube. The system includes a mandrel 
insertable into the tube, the mandrel having a pair of spaced-apart 
expandable bladders surrounding the mandrel. A first channel extends 
through the mandrel and is in communication with the bladders. A fluid 
supply circuit is in fluid communication with the first channel for 
supplying pressurized liquid to the bladders to pressurize the bladders so 
that the bladders radially expand into sealing engagement with the inner 
diameter of the tube. As the bladders expand into sealing engagement with 
the tube, a sealed annular chamber is defined between the bladders, which 
chamber is bounded by the bladders and the inner diameter of the tube. A 
second channel also extends through the mandrel, the second channel being 
in fluid communication with the fluid supply circuit at one end thereof 
and in communication with the chamber at the other and thereof in order to 
pressurize the chamber with the liquid. A controller is connected to the 
fluid supply circuit for measuring the flow rate of the liquid flowing 
through the fluid supply circuit to the chamber. Any crack in the tube 
wall adjacent the chamber will allow the liquid to escape therethrough at 
a characteristic flow rate corresponding to the size of the crack and the 
pressure in the tube. However, the fluid supply circuit maintains the 
liquid inventory and the liquid pressure in the chamber so that the liquid 
inventory and thus the pressure in the chamber remain constant as the 
liquid leaks through the crack. The controller, which is connected to a 
flow meter associated with the fluid supply circuit, measures the flow 
supply circuit, measures the flow rate and pressure of the liquid flowing 
through the fluid supply circuit. In view of the fact that the liquid 
(e.g. water) is practically incompressible, the flow rate through the 
crack and the pressure in the chamber are provided by measuring the flow 
rate and pressure of the liquid flowing through the fluid supply circuit. 
In its broad form, the present invention is a system for in situ testing of 
the leak-tightness of a tubular member, comprising: (a) a body insertable 
into the tubular member; (b) seal means surrounding said body for 
sealingly engaging the tubular member to define a sealed chamber in the 
tubular member; (c) fluid supply means in fluid communication with the 
chamber for supplying a fluid to the chamber to pressurize the chamber to 
a predetermined pressure; and (d) control means connected to said fluid 
supply means for controlling said fluid supply means, so that the 
predetermined pressure is maintained in the chamber as a breach in the 
tubular member allows the fluid to leak from the chamber. 
In its broad form, the present invention is also a method of testing the 
leak-tightness of a tubular member, comprising the steps of: (a) inserting 
a body into the tubular member; (b) defining a sealed chamber in the 
tubular member by engaging the tubular member with a pair of spaced-apart 
seals surrounding the body; (c) pressurizing the chamber by supplying a 
fluid to the chamber from a fluid supply reservoir in fluid communication 
with the chamber; and (d) maintaining a predetermined pressure in the 
chamber as a breach in the tubular member allows the fluid to leak from 
the chamber by controllably operating a controller connected to the fluid 
supply reservoir. 
An object of the present invention is to provide a system and method for in 
situ testing of the leak-tightness of a tubular member, which tubular 
member may be a heat transfer tube of the kind found in typical nuclear 
steam generators. 
Another object of the present invention is to provide a system and method 
for in situ testing of the leak-tightness of a nuclear steam generator 
heat transfer tube to obtain empirical data usable for satisfying the 
requirements of Draft Regulatory guide 1.121 (NUREG-1477). 
A feature of the present invention is the provision of a mandrel insertable 
into the tube and spaced-apart expandable bladders surrounding the 
mandrel, the bladders capable of radially expanding into sealing 
engagement with the inner diameter of the tube to define a sealed annular 
chamber between the bladders. 
Another feature of the present invention is the provision of a fluid supply 
circuit in fluid communication with the bladders and with the chamber to 
hydraulically pressurize the bladders and the chamber. 
Yet another feature of the present invention is the provision of a flow 
meter connected to the fluid supply circuit for measuring the flow rate of 
the fluid flowing through the fluid supply circuit. 
An advantage of the present invention is that empirical leak-rate data 
usable for satisfying Draft Regulatory Guide 1.121 (NUREG-1477) can now be 
obtained in a precise and cost effective manner. 
These and other objects, features and advantages of the present invention 
will become apparent to those skilled in the art upon a reading of the 
following detailed description when taken in conjunction with the drawings 
wherein there is shown and described illustrative embodiments of the 
invention.

DESCRIPTION OF THE PREFERRED EMBODIMENTS 
Disclosed hereinbelow are a system and method for in situ testing of the 
leak-tightness of a tubular member, which tubular member may be a heat 
transfer tube of the kind found in typical nuclear steam generators. 
Referring to FIG. 1, there is shown a typical nuclear steam generator, 
generally referred to as 10, for generating steam. Steam generator 10 
comprises a hull 20 having an upper portion 30 and a lower portion 40. 
Disposed in hull 20 are a plurality of vertical U-shaped heat transfer 
tubes 50 that extend through a plurality of horizontal support plates 60. 
Each tube 50 has an inner diameter 70 (see FIG. 3). As shown in FIG. 1, 
disposed in lower portion 40 is a horizontal tube sheet 80 for supporting 
the ends of each tube 60. Disposed on hull 20 are a first inlet nozzle 90 
and a first outlet nozzle 100 in fluid communication with an inlet plenum 
chamber 110 and with an outlet plenum chamber 120, respectively. A 
plurality of manway holes 130 are formed through hull 20 below tube sheet 
80 for allowing access to inlet plenum chamber 110 and outlet plenum 
chamber 120. Moreover, formed through hull 20 above tube sheet 80 is a 
second inlet nozzle 140 for allowing entry of a non-radioactive secondary 
fluid (i.e., demineralized water) into hull 20. A second outlet nozzle 150 
is disposed on the top of upper portion 30 for exit of steam from steam 
generator 10. 
During operation of steam generator 10, pressurized and radioactive primary 
fluid (i.e., demineralized water) heated by a nuclear reactor core (not 
shown) enters inlet plenum chamber 110 through first inlet nozzle 90 and 
flows through tubes 50 to outlet plenum chamber 120 where the primary 
fluid exits steam generator 10 through first outlet nozzle 100. As the 
primary fluid enters inlet plenum chamber 110, the secondary fluid 
simultaneously enters second inlet nozzle 140 to ultimately surround tubes 
50. A portion of this secondary fluid vaporizes into steam which rises 
upwardly to exit steam generator 10 through second outlet nozzle 150. The 
steam is piped to a turbine-generator set (not shown) for generating 
electricity in a manner well understood in the art. Moreover, the primary 
fluid is radioactive; therefore, for safety reasons, tubes 50 are designed 
to be leak-tight, so that the radioactive primary fluid does not commingle 
with the nonradioactive secondary fluid. 
However, due to tube wall intergranular stress corrosion cracking, some of 
the tubes 70 may degrade and thus may not remain leak-tight. Therefore, 
such leaking tubes 70 may be plugged, if required, to prevent commingling 
the radioactive primary fluid with the nonradioactive secondary fluid. 
However, it has been observed that not all leaking tubes 70 require 
plugging because a small amount of tube leakage is tolerable. This is so 
because small amounts of tube leakage do not pose a safety concern in case 
of a postulated accident. In view of this, Draft Regulatory Guide 1.121 
(NUREG-1477) allows a degraded or leaking tube to be "leak-rate tested" 
for determining whether or not the tube will have an acceptable leak rate 
during such a postulated accident (e.g., main steam line break) or during 
normal operation. If the anticipated leak-rate is acceptable, then the 
leaking tube need not be plugged. According to the invention, such a 
leak-rate test can now be performed in situ in a precise and cost 
efficient manner. 
Therefore, turning now to FIG. 2, there is shown the subject matter of the 
present invention, which is a system, generally referred to as 160, for in 
situ testing of the leak-tightness of a tubular member, which tubular 
member may be heat transfer tube 50 disposed in nuclear steam generator 
10. System 160 comprises a body or mandrel, generally referred to as 170, 
insertable into tube 50 and having seal means, such as a pair of 
expandable bladders 180a/b, thereon for reasons described hereinbelow. 
Connected to mandrel 170 is a flexible protective hose 190 protectively 
surrounding two flexible fluid supply conduits 200a/b, for reasons 
described more fully hereinbelow. Hose 190 and conduits 200a/b are 
connected to fluid supply means, generally referred to as 210, for 
supplying fluid to mandrel 170, as disclosed in detail hereinbelow. 
Control means, generally referred to as 220, is connected to fluid supply 
means 210 for controlling fluid supply means 210. In addition, a hose 
driver, generally referred to as 230, includes at least one rotatable 
wheel 235 capable of engaging hose 190 for driving or translating hose 190 
and the mandrel 170 connected thereto along the longitudinal axis of tube 
50. Moreover, a support mechanism 240 is connected to mandrel 170 for 
aligning mandrel 170 coaxially with tube 50. Support mechanism 240 is also 
capable of supporting hose 190 and mandrel 170 as hose 190 and mandrel 170 
are translated in tube 50. In this regard, support mechanism 240 may be a 
ROSA (Remotely Operated Service Arm) available from the Westinghouse 
Electric Corporation located in Pittsburgh, Pa. The structure and function 
of each of these major components of system 160 are described in more 
detail hereinbelow. 
Referring to FIGS. 3 and 4, mandrel 170 comprises a generally cylindrical 
proximal end portion 250 and a generally cylindrical distal end portion 
260 integrally interconnected by a generally cylindrical intermediate 
portion 270. It will be appreciated, with reference to the several 
figures, that the terminology "proximal end portion" is defined herein to 
mean that end portion disposed nearer the bottom of outlet plenum chamber 
110 (or inlet plenum chamber 120) and that the terminology "distal end 
portion" is defined herein to mean that end portion disposed farther away 
from the bottom of outlet plenum chamber 110 (or inlet plenum chamber 
120). Distal end portion 250 of mandrel 170 includes a generally 
cylindrical first central body 280 having an externally threaded distal 
end portion 290 and an externally threaded proximal end portion 300. 
Threadably connected to distal end portion 290 is a generally conical nose 
member 310 for easily inserting mandrel 170 into tube 50. Nose member 310 
has a step bore 320 defining an unthreaded portion 330 therein for reasons 
disclosed presently. Step bore 320 also has an internally threaded portion 
340 of smaller diameter than unthreaded portion 330 for threadably 
engaging the external threads of distal end portion 290 belonging to first 
central body 280. In this manner, nose member 310 is threadably connected 
to first central body 280. In addition, threadably connected to proximal 
end portion 300 of first central body 280 is the intermediate portion 270. 
A distal end portion 345 of intermediate portion 270 has a step bore 350 
defining an unthreaded portion 360 therein for reasons disclosed 
presently. Step bore 350 also has an internally threaded portion 362 of 
smaller diameter than unthreaded portion 360 for threadably engaging the 
external threads of proximal end portion 300 which belongs to first 
central body 280. In this manner, intermediate portion 270 is threadably 
connected to first central body 280. Moreover, a proximal end portion 365 
of intermediate portion 270 has a step bore 370 defining an unthreaded 
portion 380 therein for reasons disclosed presently. Step bore 370 also 
has an internally threaded portion 390 of smaller diameter than unthreaded 
portion 380 for threadably engaging the external threads of a proximal end 
portion 400 belonging to a generally cylindrical second central body 410. 
In this manner, intermediate portion 270 is threadably connectable to 
second central body 410. In addition, threadably connected to a proximal 
end portion 420 second central body 410 is a connector 430 having a step 
bore 440 defining an unthreaded portion 450 therein for reasons disclosed 
presently. Step bore 440 also has an internally threaded portion of 
smaller diameter than unthreaded portion 450 for threadably engaging the 
external threads of proximal end portion 420 which belongs to second 
central body 410. In this manner, connector 430 is threadably connected to 
second central body 410. A proximal end portion 460 of connector 430 may 
have a plurality of serrations 465 around the exterior surface thereof for 
engaging the inside surface of hose 190 in order to secure hose 190 to 
connector 430. 
As best seen in FIGS. 4 and 4A, a first channel or flow passage 470 extends 
longitudinally through connector 430, second central body 410, 
intermediate portion 270 and first central body 280. First channel 470 is 
in communication with a first port 480 on the exterior surface of first 
central body 280 for conducting an incompressible fluid (e.g., water, oil 
or the like) to the exterior surface of first central body 280. First 
channel 470 is also in communication with a second port 490 on the 
exterior surface of second central body 410 for conducting the 
incompressible fluid to the exterior surface of second central body 410. 
As disclosed in more detail hereinbelow, first channel 470 is in fluid 
communication with fluid supply conduit 200a for supplying the fluid to 
first channel 470. Moreover, a second channel or flow passage 494 extends 
longitudinally through connector 430, second central body 410 and into 
intermediate portion 270. Second channel 494 is in communication with a 
third port 496 on the exterior surface of intermediate portion 270 for 
conducting the fluid to the exterior surface of intermediate portion 270. 
As disclosed in more detail hereinbelow, second channel 494 is also in 
fluid communication with fluid supply conduit 200b for supplying the fluid 
to second channel 494. 
Still referring to FIGS. 4 and 4A, surrounding first central body 280 is 
the previously mentioned seal means or generally tubular first bladder 
180a which maybe formed from a resilient thermo elastomer material. First 
bladder 180a has an inner surface 500 that covers first port 480. One end 
of first bladder 180a is disposed in unthreaded portion 330 of nose member 
310. This end of first bladder 180a is sized to be tightly sealingly 
interposed between first central body 280 and nose member 310, such 
sealing being preferably obtained by a press fit. The other end of first 
bladder 180a is disposed in unthreaded portion 360 of intermediate portion 
270. This end of first bladder 180a is sized to be tightly sealingly 
interposed between first central body 280 and intermediate portion 270, 
such sealing being preferably obtained by a press fit. Thus, first bladder 
180a is sealingly connected to first central body 280 so that the 
interface between first central body 280 and the opposite ends of first 
bladder 180a is leak-tight as first bladder 180a is expansively 
pressurized, in the manner described hereinbelow. 
Referring again to FIGS. 4 and 4A, surrounding second central body 410 is 
the previously mentioned seal means or generally tubular second bladder 
180b which also may be formed from the resilient thermo elastomer material 
mentioned hereinabove. Second bladder 180b has an inside surface 510 that 
covers second port 490. One end of second bladder 180b is disposed in 
unthreaded portion 380 of intermediate portion 270. This end of second 
bladder 180b is sized to be tightly sealingly interposed between second 
central body 410 and intermediate portion 270, such sealing being 
preferably obtained by a press fit. The other end of second bladder 180b 
is in unthreaded portion 450 of connector 430. This end of second bladder 
180b is sized to be tightly sealingly interposed between second central 
body 410 and connector 430, such sealing being preferably obtained by a 
press fit. Thus, second bladder 180b is sealingly connected to second 
central body 410, so that the interface between second central body 410 
and the opposite ends of second bladder 180b is leak-tight as second 
bladder 180b is expansively pressurized in the manner described 
hereinbelow. As bladders 180a/b are pressurized to radially expand and 
intimately engage inner diameter 70 of tube 50, they will define a sealed 
annular chamber 515 therebetween (see FIG. 4A), the purpose of bladders 
180a/b is provided hereinbelow. 
Referring now to FIG. 5, system 160 further comprises the previously 
mentioned fluid supply means 210 which supplies fluid to mandrel 170. In 
this regard, fluid supply means 210 includes a pressurized gas fluid 
reservoir, such as a compressed air supply 520, connected to a suitable 
filter 530, such as by a pipe 540. Filter 530 filters the air stream 
passing through pipe 540 in order to remove any dirt water and foreign 
particulate matter therefrom. The purpose of air supply 520 is to operate 
an air pump 550, for reasons disclosed hereinbelow. It is important that 
filter 530 remove the dirt, water and foreign particulates from the air 
stream in order to sustain the operating life of air pump 550. Sustaining 
the operating life of air pump 550 ensures that the pressure of the fluid 
exiting pump 550 is repeatable. Air operated pump 550 has a muffler device 
555 associated therewith for exhausting the air supplied to air operated 
pump 550. Filter 530 is connected, such as by a pipe 560, to an air 
regulator 570. The filtered air flows from filter 530, through pipe 560 
and into air regulator 570. From air regulator 570, the air flows into a 
pipe 580 interconnecting air regulator 570 and a first valve 590. Air 
regulator 570 is adapted to regulate or control the pressure of the air 
passing into pipe 580. First valve 590 may be a solenoid valve. A pressure 
gauge 600 may be in fluid communication with pipe 580, as at location 610, 
for displaying the air pressure in pipe 580 to the operator of system 160. 
A pipe 620 interconnects first valve 590 with air operated pump 550. 
Still referring to FIG. 5, air operated pump 550 is capable of pumping an 
incompressible fluid or liquid (e.g., water, oil or the like) from a 
liquid reservoir 630 which is connected to it, such as by pipe 640. Pump 
550 is itself connected, such as by a pipe 650, to what is termed herein a 
"dual flow circuit", generally referred to as 660 Pipe 650 is connected to 
dual flow circuit 660 as at location 670. Dual flow circuit 660 includes 
the previously mentioned first fluid supply conduit 200a, which first 
conduit 200a supplies the liquid to first channel 470. Dual flow circuit 
660 also includes the previously mentioned second fluid supply conduit 
200b, which second conduit 200b supplies the liquid to second channel 494. 
Disposed in first conduit 200a is a second valve 680, which may be a 
solenoid valve. Connected to first conduit 200a, as at location 690 
downstream of second valve 680, is a first pressure transducer 695 for 
sensing the liquid pressure in first conduit 200a as the liquid flows 
through first conduit 200a. In addition, disposed in second conduit 200b 
is a third valve 700, which also may be a solenoid valve. That is, third 
valve 700 is capable of adjusting the pressure of the liquid flowing in 
second conduit 200b in order to adjust or ramp the pressure in chamber 515 
to a predetermined value. Connected to second conduit 200b, as at location 
710 downstream of third valve 700, is a flow meter 720 for measuring the 
flow of the liquid in second conduit 200b. Moreover, connected to second 
conduit 200b, as at location 730 downstream of flow meter 720, is a second 
pressure transducer 740 for sensing the pressure of the liquid in second 
conduit 280b as the liquid flows through second conduit 280b. In addition, 
interconnecting second pressure transducer 740 and third valve 700 may be 
a "feedback loop" in the form of an electrical conductor wire 750 for 
controlling the operation of third valve 700 in order to control the flow 
rate (and thus the pressure) of the liquid flowing in second conduit 200b. 
In addition, interconnecting second pressure transducer 740 and third 
valve 700 may be a "feedback loop" in the form of an electrical conductor 
wire 750. In this regard, the following will describe the manner in which 
pressure transducer 740 interfaces or cooperates with third valve 700. 
Third valve 700, which controls the pressure, provides an output pressure 
proportional to its electrical command signal input. Furthermore, 
interconnecting pipe 650 and an accumulator tank 760, as at location 770, 
is a pipe 780 having a fourth valve 790 disposed therein for controlling 
the flow of liquid to and from accumulator tank 760. Fourth valve 790 may 
be a solenoid valve. As shown in FIG. 5, accumulator tank 760 has a 
predetermined amount of liquid therein for maintaining the liquid pressure 
in pipe 650 and thus for maintaining the liquid pressure in dual fluid 
supply circuit 660, which is connected to pipe 650. The presence of 
accumulator tank 760 will dampen inadvertent pressure spikes in fluid 
supply means 210. More specifically, the purpose of accumulator tank 760 
is to remove or smooth-out pressure spikes or pulsations produced by the 
double-acting reciprocating air pump 550. The presence of accumulator tank 
760 allows for accurate control and measurement of the pressure and fluid 
flow in fluid supply means 210. 
Referring yet again to FIG. 5, system 160 further comprises the previously 
mentioned control means 220 electrically connected to fluid supply means 
210 for controlling fluid supply means 210. Control means 220 includes a 
controller 800 electrically connected, such as by an electrical conductor 
wire 810, to first pressure transducer 695 for measuring and recording the 
liquid pressure sensed by first pressure transducer 695. Controller 800 is 
also electrically connected, such as by an electrical conductor wire 820, 
to flow meter 720 for measuring and recording the flow rate of the liquid 
metered by flow meter 720. Moreover, controller 800 is electrically 
connected, such as an by electrical conductor wire 830, to second pressure 
transducer 740 for measuring and recording the pressure sensed by second 
pressure transducer 740. Controller 800 includes display means, such as a 
paper strip chart 840, for displaying the flow rate and pressures measured 
and recorded by controller 800. Controller 800 may be computerized for 
automatically measuring and recording the flow rate and pressures and for 
automatically controlling fluid supply means 210. 
OPERATION 
The operation of system 160 will be described with respect to establishing 
the previously mentioned database. 
In this regard, steam generator 10 is first removed from service in a 
manner well understood in the art and system 160 is transported 
sufficiently near steam generator 10 to perform the leak test. 
In this regard, hose driver 230 is connected to manway 130 and support 
mechanism 240 is installed in outlet plenum chamber 110 (or inlet plenum 
chamber 120). The hose driver 230 and support mechanism 240 are used in 
the customary manner to translate an eddy current probe (not shown) in 
tube 50 for detecting the location of the degraded portion (not shown) of 
tube 50. As the eddy current probe encounters the degraded portion of tube 
50, it will generate a voltage amplitude uniquely characteristic of the 
degradation. This voltage amplitude is suitably recorded, such as in a 
data base. The eddy current probe is then withdrawn from tube 50. 
Next, mandrel 170 is inserted through manway 130 and into outlet plenum 
chamber 110 (or into inlet plenum chamber 120), whereupon it is engaged by 
support mechanism 240 for aligning the longitudinal axis of mandrel 170 
with the longitudinal axis of tube 50. Wheels 235 of hose driver 230 are 
caused to engage hose 190 and rotate so that mandrel 170 advances to the 
location of the degradation in tube 50 identified by the previously 
mentioned eddy current probe. In this regard, mandrel 170 is advanced in 
tube 50 such that intermediate portion 270 is positioned adjacent the 
degradation. 
Next, fluid supply means 210 is operated to radially expand bladders 180a/b 
into intimate sealing engagement with inner diameter 70 of tube 50, such 
that annular sealed chamber 515 is defined therebetween. More 
specifically, first valve 590, second valve 680 and fourth valve 790 are 
opened as third valve 700 is closed. Compressed air supply 520 is then 
operated to supply compressed or pressurized air into pipe 540, through 
filter 530, through pipe 560, through air regulator 570, through open 
first valve 590 and to air operated pump 550. As air operated pump 550 
operates, it withdraws the water from fluid reservoir 630, which water 
flows through pipes 640 and 650 to be received by dual flow circuit 660. 
The water then flows through first conduit 200a, through open second valve 
680 and to first channel 470 which is formed in mandrel 170. After passing 
into first channel 470, the water will exit first and second ports 
480a/490 to pressurize bladders 180a/b, respectively, so that bladders 
180a/b sealingly engage inner diameter 70 to define sealed annular chamber 
515 therebetween. 
After bladders 180a/b are thusly pressurized to a first predetermined 
pressure (e.g., approximately 4,000 psia), second valve 680 is closed and 
third valve 700 is opened to precisely pressurize chamber 515 to a second 
predetermined pressure, which second predetermined pressure may be the 
anticipated pressure during normal operation or during a main steam line 
break accident (e.g., approximately 2,750 psia). Third valve 700 is 
capable of adjusting or ramping the pressure in second conduit 200b so 
that any number of second predetermined pressures are obtainable. As the 
water is pumped to dual flow circuit 660, it will flow through second 
conduit 200b, through open third valve 700 and to second channel 494 in 
mandrel 170. 
As chamber 494 is pressurized, the pressurized water therein will leak, 
seep or flow out chamber 494 through any through-wall crack (not shown) 
adjacent chamber 494. However, as the water flows through the crack, the 
inventory of water in chamber 494 will remain constant because fluid 
supply means 210 will continue to supply water to chamber 494 to maintain 
the predetermined pressure therein. Moreover, because the water inventory 
in chamber 494 remains constant, the flow rate of the water supplied 
through second fluid supply conduit 200b must necessarily exactly equal 
the flow rate of the water flowing through the crack in the degraded 
portion of tube 50. Thus, it will be appreciated from the description 
hereinabove, that measuring the flow rate of the water flowing through 
second conduit 200b will simultaneously exactly measure the flow rate of 
the water flowing through the crack. It will also be appreciated from the 
description hereinabove, that the pressure in chamber 515 must necessarily 
exactly equal the pressure in second conduit 200b. 
The flow rate of the water flowing through second conduit 200b is measured 
and recorded by controller 800, which is electrically connected to flow 
meter 720 that is in turn hydraulically connected to second conduit 200b. 
Moreover, the pressure in chamber 515 is also measured and recorded by 
controller 800, which is electrically connected to second pressure 
transducer 740 that is in turn hydraulically connected to conduit 200b. In 
this manner, the flow rate of the water through the crack and the pressure 
in chamber 555 are precisely measured by controller 800. Controller 800 
also displays and records, such as by the paper strip chart 840, the flow 
rate through the crack and the pressure in chamber 515. 
Thus, it will be understood that three empirical data values are obtainable 
characterizing the crack in tube 50. First, the eddy current measurement 
provides a voltage amplitude value uniquely associated with the crack. 
Second, at least one value of pressure (i.e., the predetermined pressure 
in chamber 515) has been obtained. Third, at least one value of the flow 
rate of the water through the crack at the pressure in chamber 515 has 
been obtained. Of course, a plurality of pressure values and associated 
flow rates corresponding to each of a plurality the voltage amplitude 
readings are preferably measured and recorded in order to establish a 
suitable data base. That is, an eddy current voltage amplitude measurement 
for each of a plurality of degraded portions of the same tube or different 
tubes may be taken and the leak-tightness testing process repeated for 
each voltage amplitude reading. 
Such a data base may be used to conveniently determine the anticipated flow 
rate at a given pressure through a crack in a heat exchanger tube 
belonging to an operating nuclear steam generator simply by performing an 
eddy current measurement on the tube. That is, an eddy current measurement 
is performed on the tube to obtain a voltage amplitude value uniquely 
associated with a degraded portion thereof. Next, that voltage amplitude 
value and the anticipated tube pressure expected during normal operation 
or during a postulated accident (e.g., main steam line break) are 
looked-up or found in the previously established data base. The data base 
will therefore provide the flow rate through the crack corresponding to 
the measured voltage amplitude and postulated pressure. 
It will be appreciated from the description hereinabove, that an advantage 
of the present invention is that the empirical leak-rate data for 
satisfying Draft Regulatory Guide 1.121 (NUREG-1477) can now be obtained 
in a cost effective manner. This is so because the not inconsiderable 
expense associated with constructing a bench-scale model steam generator 
for testing is avoided because the testing is performed on full-sized heat 
transfer tubes. In addition, the expense associated with removing leaking 
full-sized steam generator tubes for testing is also avoided because, by 
use of the invention, such leaking tubes are examined in situ. 
It will also be appreciated from the description hereinabove, that another 
advantage of the present invention is that tube leak-rate data are now 
precisely obtainable. This is so because leak-rate testing is performed 
directly on an actual tube disposed in an operating full-sized steam 
generator, rather than on a bench-scale model steam generator tube which 
may or may not evince the true tube characteristics existing in an 
operating full-sized steam generator. Also, the data obtained by use of 
the invention are more precise for yet another reason. That is, if tubes 
are removed from a full-sized steam generator for testing, there is a risk 
that the morphology of the degraded portion of the tube will be different 
after removal than before removal. This leads to imprecise tube testing 
data. By use of the invention, the risk of change in tube morphology is 
avoided because the tube is tested in situ. 
Although the invention is illustrated and described herein in its preferred 
embodiment, it is not intended that the invention as illustrated and 
described be limited to the details shown, because various modifications 
may be obtained with respect to the invention without departing from the 
spirit of the invention or the scope of equivalents thereof. For example, 
a suitable eddy current coil may be integrally attached to mandrel 170 for 
obtaining the required voltage amplitude, rather than using a separate 
inspection instrument for performing the eddy current measurement. 
Moreover, although system 160 is described as suitable for leak-rate 
testing a nuclear steam generator heat transfer tube, system 160 is also 
usable whenever in situ leak-rate testing of a tubular member is desired. 
Therefore, what is provided are a system and method for in situ testing of 
the leak-tightness of a tubular member, which tubular member may be a heat 
transfer tube of the kind found in typical nuclear steam generators.