ML19309A272
| ML19309A272 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/01/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309A267 | List: |
| References | |
| NUDOCS 8003270603 | |
| Download: ML19309A272 (9) | |
Text
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- REACER CCCLANT SYSTBL STE.41 GENERATORS LBIITING CONDITICN FOR CPE?ATION 3.4.5.1 Primary-to-seccndarf leakage through the steam generator tubes shall be li i ed to 1 GB1 total for all steam generators.
t 3.4.5.2 Each steam generator shall be CPERA3LE with a water level between
( ) and ( ) inches.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With any steam generator tube leakage greater than the above limit reduce the leakage rate within four hours or be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
b.
With one or more steam generators inoperable due to steam generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing T above 2000F.
WithoneormoresteamgeneratorsinoperabSeduetothewater a
c.
level being outside the limits, 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 /> and in COLD SHUTC0WN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE r.EOUIREMENTS
,4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance
,of the following augmented inservice inspection program and the requirements
! of Specification 4.0.5.
il4.4.5.1.. Steam Gene'rator Samole Selection and Inscection - Each steam generator shall be ceterminec OPERABLE curing snutcown oy selecting and
- inspecting at least the minimum number of steam generators specified in Table 4.4-1.
'4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam generator tuce minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspecticn of steam generator tubes shall be performed at the frequencies specified in Specification 4. A.5.3 and the inspected tubes shall
~ he verified acceptable per the acceptance criteria of Specification 1.4.5.1
$he tubes selected for each inservice inspection shall include at least 35 bf the total number of tubes in all steam generators; the tubes selected '
'for these inspecticns shall be selected on a randem basis except:
Where experience in similar plants with similar water chemistry a.
indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.
8003 270 $O3 B&W-STS 3/4 4-6 June 1, 1975
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued).
b.
The first inservice inspection (subsequent to the preservice inspection) of each steam generator,shall include:
1, All nonplugged tubes that previously had detectable wall penetrations (.>20%),and 2,
Tubes in those areas where experience has indicated poten-ttal problems.
c',
The second and third inserYice inspections may be less tcAn a full tube inspection by concentrating (selecting at least 50%
of the tubes to be inspectedl the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category Insoection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C,4 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhttit significant.(>10%) further wall penetrations to be included in the above percentage calculations.
4,4,5,3 Inscection Frecuencies - The above required inservice inspections I
of steam generator tuces snail be perfomed at the following frequencies:
a, The first inservice inspection shall bc ::erfor=ed after 5 Effective Full Power F.onths but within 2a calendar months of initial criticality. Subsequent inservice inspections shall be B W-STS 3/4 4 7 June 1,19J6
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecu-tive inspections fcilowing service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two cor.secutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
'b.
If the inservice inspection of a steam generator ccnducted in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.
c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.
1.
Primary-to-secondary tubes leaks (not including leaks originating from tube-to tube sheet welds) in excess of the limits of Specification 3.4.5.1, 2.
A seismic occurrence greater than the Operating Basis Earthquake, 3.
A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.
A main steam line or feedwater line break.
4.4.5.4 Acceotance Criteria a.
As used in this Specification:
1.
Imoerfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications Belcw 20% of the nominal tube wall thickness, if de:ectable, may be considered as imperfections.
B&W-STS 3/4 4-8 Jane 1, 1976
' REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
3.
Degraded Tube means a tube containing imperfections >20%
cf the nominal wall thickness caused by degradation. _
4.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceecs the plugging limit. A tube containing a defect is defective. Any tube which does not pemit the passage of the eddy-current inspection probe shall be deemed a defective tube.
6.
pluqqing Limit means the imperfection depth at or beyond wnien tne tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the ncminal tube wall thickness.
7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.
8.
Tube Insoection means an inspection of the steam generator tube frem tne point of entry ccmpletely to the point of exi t, b.
The steam generator shall be detemined OPERABLE after ccmpleting the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2 4.4.5.5 Recorts a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 15 days.
b.
The ccmplete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was conpleted. This report shall include:
B&W-STS 3/4 4-9 June 1, 1976
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged.
Results of steam generator tube inspections which fall into c.
Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
4.4.5.6 The steam generator shall be demonstrated OPERABLE by verifying steam generator level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B&W-STS 3/4 4-10 June 1, 1976 9
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W B&W-STS 3/4 4-11 June 1, 1976
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TABLE 4.4-2 o,
ist STEAM GENERATOft TUDE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPl.E INSPECTION Sample Site flesult Action flequired flesult Action flequired flesult Action Hequired A rninimum of C-1 None N/A N/A N/A N/A S Tubes per S. G.
C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S subes in this S. G.
C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G.
Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample
.t**
C-3 Inspect all tubes in All other
- b this S. G., plug de-S. G.s are None N/A N/A lective tubes and C-1 4
inspect 2S tubes in Perform action for N/A N/A rY each other S. G.
_[g tn C-2 result of second eukhtional sample Prompt notification S. G. are to NilC gmsuant C-3 to specification Additional inspect all tubes in 6.0.1 S. G. is C-3 each S. G. and plug defective tubes.
Prompt notification N/A N/A to NilC pursuant to specification L
6.9.1 C
s 3
(D S = 3 b Where i is the nurnber of steam generators in the unit, and n is the number of steam generators inspected l
n during an inspection 1
-o cn 4
k1
REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.
The low level limit is based on providing enough water volume to prevent a pressurizer low level or a reactor coolant system low pressure condition that would actuate the Reactor protection System or the Engineered Safety Feature Actuation System as a result of a reactor scram. The high level limit is based on providing enough steam volume to prevent a pressurizer high level as a result of any transient.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. 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 surveillance 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.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If tne secondary coolant chemistry is not maintained within these chemistry limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be 'imited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 GpM).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads B&W-STS B 3/4 4-2 June 1, 1976
REACTOR COOLANT SYSTEM BASES imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry the treatment of the secondary coolant. However, even if a defect should develop in service, 'c will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspec-tion fall into Category C-3, these results will be promptly reported to the Ccmmission pursuant to Specification 6.9.1 prior to resumption of plant operation. Such cases will be considered by the Ccmmission on a case-by-case basis and may result in a requirement for analysis, labora-tory examinations, tests, additional edd-current inspection, and re-vision of the Technical Specifications, if necessary.
The steam generator water level limits are consistent with the initial assumptions in the FSAR.
B&W-STS B 3/4 4-3 June 1, 1976
_