ML20055H006
| ML20055H006 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/27/1990 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20055H007 | List: |
| References | |
| NUDOCS 9007250043 | |
| Download: ML20055H006 (10) | |
Text
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2.1 SAFETY LIMITS RASES 2.1.1 REACTOR CORE The restrictions O this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive claddin from nucleate boiling (DNB)g temperatures because of the onset of departureand the r coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-1 correlation.
The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuhiform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.
The minimum value of DNBR during steady state, normal o)erational transients, and anticipated transients is limited to 1.17, tie DNBR design limit of the WRB-1 correlation.
This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur.
Margin is maintained by meeting a DNBR value of 1.27 in the safety analyses.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER Reactor Coolant System pressure and average temperature for which the minimum ONBR is no less than 1.27, or the average enthalpy at the vessel exit is equal l
to the enthalpy of saturated liquid.
Thesecurvesarebasedonanenthalpyhotchannelfactor,FfH,of1.52and a reference cosine with a peak of 1.61 for axial power shape.
An allowance is included for an increase in Ffg at reduced power based an the expression:
Ffg=1.52[1+0.3(1-P))
l Where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f3 (AI) function of the Overtemperature trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-L temperature AT trips will reduce tha Setpoints to provide protection consistent with core Safety Limits.
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PDC SOUTH TEXAS - UNITS 1 & 2 B 2-1 June 27, 1990 i
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o 2.2 LIMITING SAFETY SYSTEM SETTINGS RA m 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limiu specified in Table 2.2-1 are the nominal values at which the Reactor trips era set for each functional L. nit.
The Trip Setpoints have been selected to ensura that the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents.
The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as-measured" Setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occw between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.
Operation with Setpoints less conservative than the Trip Set-point but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combi-nation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equa-tion 2.2-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, and the "as-measured" values of the errors are considered.
Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.
TA or Total Allowance is the difference, in percent span, bet 1een the Trip Setpoint and the value used in the analysis for Reactor trip.
R or Rack Error is the "as-measured" devia-tion, in percent span, for the affected channel from the specified Trip Set-point.
S G Sensor Error is either the "as-measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions.
Use of Equation 2.2-1 allows for a sensor drift factor and an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Valua exhibits the behavior that the rack has not met its allowance.
Because there is a small statistical chance that this will happen, an infrequent excessive drif t is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
I 1
SOUTH TEXAS - UNITS 1 & 2 B 2-3
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i 2.1 SAFETY LIMITS gs 2.1.1 REACTOR CORE l
The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission oroducts to the reactor coolant.
Overhecting of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the i
heat transfer coefficient is large and the cladding surface temperature is l
slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-1 correlation.
The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.
The minimum value of DNBR during steady state, normal o)erational transients, and anticipated transients is limited to 1.17, tie DNBR design j
limit of the WRB-1 correlation.
This value corresponds to a 95% probability J
at a 9% confidence level that DNB will not occur.
Margin is maintained by j
meet'og a DNBR value of 1.27 in the safety analyses.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER I
Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.27, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
Thesecurvesarebasedonanenthalpyhotchannelfactor,FfH,of1.52and a reference cosine with a peak of 1.61 for axial power shape.
An allowance is i
included for an increase in Ffg at reduced power based on the expression:
Ffg=1.5?[1+0.3(1-P)]
Where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip.
When the axial power imbalance 3is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Saftty Limits.
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PDC SOUTH TEXAS - UNITS 1 & 2 B 2-1 June 27, 1990
s SAFETY LIMITS aAsrs 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby. prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which pemits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
demonstrate integrity prior to initial operation.psig) of design pressure, to The entire RCS is hydrotested at 125% (3110 SOUTH TEXAS
'INITS 1 & 2 B 2-2
2.2 LIMITING SAFETY SYSTEM SETTINGS RASES i
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functions 1 unit.
The Trip Setpoints have been selected to ensure that the core and Reactor Coolant
]
System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents.
The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as-measured" Setpoint is within the band allowed for calibration accuracy.
1 To accommodate the instrument drift assumed to occur between operational j
tests and the accuracy to which Setpoints can be measured and calibrated.
Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.
Operation with Setpointe less conservative than the Trip Set-point but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
An optional provision has been included for determining the OPERABILITY of a channel when its Trip l
Setpoint is found to exceed the Allowable Value.
The methodology of this option utilizes the "as measured" deviation from the specified calibration l
point for rack and sensor components in conjunction with a statistical combi-nation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equa-tion 2.2-1, Z + R + S < TA, the interactive effectr of the errors in the rack and the sensor, and the "as-measured" values of the errors are considered.
Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and l
rack drift and the accuracy of their measurement.
TA or Total Allowance is j
the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip.
R or Rack Error is the "as-measured" devia-tion, in percent span, for the affected channel from the specified Trip Set-point.
S or Sensor Error is either the "as-measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent i
span, from the analysis assumptions.
Use of Equation 2.2-1 allows for A sensor drif t factor and an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Because there is a small statistical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
SOUTH TEXAS - UNITS 1 & 2 B 2-3
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LIMITING SAFETY SYSTEM SETTINGS
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aA1F1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
.The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.
In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System.
The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated.
This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary j
actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.
Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting.
The Low Setpoint trip provides protection during suberitical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from e?1 power levels.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this tr h complements the Power Range Neutron Flux High and Low l
trips to ensure that the criteria are met for rod ejection from mid power.
The Power Range Negative Rate trip provides protection for control rod drop accidents.
At high power a single or multiple rod drop accident could cause i
local flux peakina which cauld cause an unconservative local DNBR to exist.
The Power Range Negatlve Rate trip will prevent this from occurring by tripping the reactor.
No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the design limit.
SOUTH TEXAS - UNITS 1 & 2 B 2-4 June 27, 1990
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- Ms INDICATED AXIAL Pl.UX DIFFERENCE FIGURE B 3/4.2-1 i
TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER i
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-3
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POWER DISTRIBUTION LIMITS RARES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL-FAGIUR (Continued) c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
FfH will be maintained within its limits provided Conditions a. through
- d. above are maintained.
The combination of the RCS flow requirement (395,000 gpm)andtherequirementonFfH guarantees that the DNBR used in the safety analysis will be met.
TherelaxationofFfHasafunctionofTHERMAL i
POWER allows changes in the radial power shape for all permissible rod inser-tion limits.
WhenFfH is measured, no additional allowances are necessary prior to comparison with the limit.
Ameasurementerrorof4%forFhhasbeenallowed for in the determination of the design DNBR value.
Fuel rod bowing reduces the value of DNB ratio.
Margin has been maintained between the DNBR value used in the safety analyses (1.27) and the design limit (1.17) to offset the rod bow penalty and other penalties which may apply.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-4 June 27, 1990
o 4
POWER DISTRIBUTION LIMITS I
RASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) q easurement is taken, an allowance for both experimental error When an F m
and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance, t
i l
The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit.
The i
9 F
limit for RATED THERMAL POWER (F RTP) as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.6 was determined from expected l
power control manuevers over the full range of burnup conditions in the core, l
l 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-l tion satisfies the design values used in the power capability analysis, i
Radial power distribution measuretents are made during STARTUP testing and l
periodically during power operation, The limit of 1.02, at which corrective action is required, provides DNB l
and linear heat generation rate protection with x-y plane power tilts.
A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identification and carrection of a dropped or nisaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum allowed q
power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the l
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5
I-l POWER DISTRIBUTION LIMITS RASFS 3/4.2.5 DNB PARAMETERS (Continued) initial FSAR a umptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the design limit throughout each analyzed transient.
The indicated T value of 598'F and the indicated pressurizer pressure value of 2201 psig aN9provided assuming that the readings from four channels will be averaged before comparing with the required limit.
The flow requirement (395,000 gpa) includes a measurement uncertainty of 3.5%.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
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l SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-6 June 27, 1990
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3/4.4 REACTOR COOLANT SYSTEM l
RASES l
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION l
The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal opera-tions and anticipated transients.
In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufttcient heat removal i
capability for removing core decay heat even in the event of a bank withdrawal accident; however a single reactor coolant loop provides sufficient heat removalcapacityIfabankwithdrawalaccidentcanbeprevented,i.e.,by opening the Reactor Trip System breakers.
Single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal ca nbility for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure l
considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
l-The boron dilution analysis assumed a common RCS volume, and maximum di-lution flow rate for MODES 3 and 4, and a different volume and flow rate for MODE 5.
The MODE 5 conditions assumed limited mixing in the RCS and cooling with the RHR system only.
In MODES 3 and 4, it was assumed that at least one reactor coolant pump was operating.
If at least one reactor coolant pump is not operating in MODE 3 or 4, then the maximum possible dilution flow rate must be limited to the value assumed for MODE 5.
The operation of one reactor coolant pump (RCP) or one RHR pump provides l
adequate flow to ensure mixing, prevent stratification and produce gradual I
reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting an RCP with one or more RCS cold legs less i
than or equal to 350'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Apperdix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricthg startin of the RCPs to when the secondary water temperature of each l
steam gentrator is ess than 50'F above each of tM dCS cold leg temperatures.
3/4.4.2 SAFETY VALW S.
The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 504,950 lbs per hour of saturated steam at the valve setpoint of 2500 psia.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the l
SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-1 June 27, 1990 1
. ~
, REACTOR COOLANT SYSTEM RASFS SAFETYVALVES(Continued) zatIon,ovidesoverpressurereliefcapabilityandwillpreventRCSoverpressuri-RCS pr In addition, the Overpressure Protection System provides a diverse I
means of protection against RCS overpressurization at low temperatures.
During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
i j
The combined relief capacity of all of these valves is greater than the maximum i
surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the turbine trip resulting from loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during i
shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the parame-ter is restored to within its limit following expected transient operation.
The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraalically solid system.
The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and estabi hh natural circulation.
3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients ur to and including the design step load decrease with steam dump.
Operati o of the PORVs minimizes I
the undesirable opening of the spring-loaded pressurizer Code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
l 3/4.4.5 STEAM GENERATORS L
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2
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