ML20210D297
| ML20210D297 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/04/1987 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML19292H207 | List: |
| References | |
| TAC-65287, NUDOCS 8705070111 | |
| Download: ML20210D297 (17) | |
Text
ENCLOSURE 2 Proposed Changed Pages Unit 2 Revision Page 3/4 4-9 Replace Page 3/4 4-10 Replace Page 3/4 4-12 Replace Page 3/4 4-13 Repl ace Page 3/4 4-13a Add Page 3/4 4-15 Repl ace Page B3/4 4-3a Replace
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REACTOR COOLANT SYSTEM 3/4.4.6 STEAM GENERATORS LIMITING CONDITION FOR OPERATION
=
3.4.6 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200 F.
SURVEILLANCE REQUIREMENTS
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4.4.6.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.S.
4.4.6.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.6.2.1 Steam Generator Tube # Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.6.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.6.4.
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators. Selection of tubes to be inspected is not affected by the F* designation. When applying the exceptions of 4.4.6.2.1.a through 4.4.6.2.1.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection.
The tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.
b.
The first sample of tubes selected for each inservice inspection l
(subsequent to the preservice inspection) of each steam generator shall l
include:
1 1.
All nonplugged tubes that previously had detectable wall penetrations greater than 20%.
- When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.6.4.a.9.
FARLEY-UNIT 2 3/4 4-9 AMENDMENT NO.
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
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2.
Tubes in those areas where experience has indicated potential problems.
3.
At laast 3% of the total number of sleeved tubes in all three steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and the sleeve.
4.
A tube inspection (pursuant to Specification 4.4.6.4.a.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
c.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection 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-3 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 or sleeves must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
4.4.6.2.2 Steam Generator F* Tube Inspection _ - In addition to the minimum sample size as determined by Specification 4.4.6.2.1, all F* tubes will be inspected within the tubesheet region. The results of this inspection will not be a cause for additional inspections per Table 4.4-2.
FARLEY-UNIT 2 3/4 4-10 AMENDMENT NO.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
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4.4.6.4 Acceptance Criteria a.
As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal wall thickness, if detectable, may be considered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.
3.
Degraded Tube means a tube, including the sleeve if the tube has been repaired, that contains imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
% Degradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing a defect is defective.
6.
Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e., sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply to the area of the tubesheet region below the F* distance in F* tubes.
For a tube that has been sleeved, through wall penetration of
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greater than or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging.
7.
Unserviceable describes the condition of a tube or sleeve 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 4.4.6.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube.
9.
Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178 Rev.1, which is used to maintain a tube in service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure.
FARLEY-UNIT 2 3/4 4-12 AMENDMENT NO.
I REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 8321333322B 3333333 33 3 333 333 333 33333 33333 3333 333322 E 33 33 3 33333333 3333333 33 333333333
- 10. Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11.
F* Distance is the distance of the expanded portion of a tube
(
wnich provides a sufficient length of undegraded tube expansion to resist pullout of the tube from the tubesheet.
The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.
l 12.
F* Tube is a tube:
a) with degradation equal to or greater than 40% below the F*
distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains in service.
- 13. Tube Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all tubes exceeding the plugging or repair limit) required by Table 4.4-2.
4.4.6.5 Reports a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged, repaired or designated F* in each steam generator shall be reported to the Commission within 15 days of the completion of the inspection, plugging or repair effort.
b.
The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1.
Number and extent of tubes and sleeves inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or repaired.
FARLEY-UNIT 2 3/4 4-13 AMENDMENT NO.
REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 33333333333333333333333333333333333333333333=3333333333333333333333333333333333=3 a
c.
Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. Tne written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.
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FARLEY-UNIT 2 3/4 4-13a AMENDMENT NO.
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TABLE 4,4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required m
[
A minimum of C-1 None N/A N/A N/A N/A S Tubes per z
Z S.G.
C-2 Plug or repair C-1 None N/A N/A N
defective tubes and inspect C-2 Plug or repair C-1 None additional 2S tubes defective tubes in this S.G.
and inspect C-2 Plug or repair additional 4S tubes defective tubes in this S.G.
C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A j$
C-3 result of first sample EA C-3 Inspect all tubes All other in this S.G., plug S.G.s are None N/A N/A or repair defective C-1 tubes and inspect 2S tubes in each Some S.G.s Perform action for N/A N/A other S.G.
C-2 but no C-2 result of additional second sample S.G.s are C-3 Notification to Additional Inspect all tubes N/A N/A NRC pursuant to S.G. is in each S.G. and 10CFRSO.73 C-3 plug or repair k
defective tubes.
2 Notification to 5!
NRC pursuant to E{
10CFRSO.73 5
S = 3 N % Where N is the number of steam generators in the unit, and n is the number of steam generators Ti inspected during an inspection.
NOTE: F* tubes do not have to be plugged or repaired.
BASES
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- 3. The tube plugging limit continues to apply to the portion of the tube in the entire upper joint region and in the lower roll expansion. As noted above the sleeve plugging limit applies to these areas also.
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- 4. The tube plugging limit continues to apply to that portion of the tube above the top of the upper joint.
F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.
Included in this distance is an allowance of 0.25 inch for eddy current elevation measurement uncertainty.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20%
of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10CFR50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if 4
necessary.
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FARLEY-UNIT 2 B3/4 4-3a AMENDMENT NO.
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- _ _,,. ~. _. -, _ _. -, _, _ _ _,
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ENCLOSURE 3 l
Significant Hazards Evaluation Pursuant to 10CFR50.92 for the Proposed Steam Generator Tubesheet Region Plugging Criterion Technical Specification Change
l NS-RCSCL-86-225 Rev. 2 PAGE 2 OF 8 TUBESHEET REGION PLUGGING CRITERION J. M. FARLEY NUCLEAR PIANT UNIT 2 STEAM GENERATORS SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS INTRODUCTION As required by 10 CFR 50.91 (a) (1) this analysis is provided to demonstrate that a
proposed license acendment to implement alternate tubesheet region tube plugging criteria for the J. M.
Farley Unit 2 steam generators represents no significant hazards consideration.
In accordance with the three factor test of
^
implementation of the proposed license amendment was analyzed using the following standards and found not to:
- 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a
significant reduction in a margin of safety.
The amendment has been proposed due to addy current indications i
of tube degradation in the roll expanded portion of the tubes within the tubesheet in the steam generators at Farley Unit 2.,
These steam generators were fabricated with a mechanical roll expansion of the tube the full depth of the tubesheet.
It has been determined through interpretation of addy current l
examinations, that the tube degradation occurring in the Farley steam generators is of the type associated with primary water stress corrosion cracking (PWSCC).
Using existing Technical Specification tube plugging
- criteria, many of the tubes with these indications would have to be removed from service.
It can be shown that tube plugging or repair is not required in many cases to maintain tube bundle integrity.
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NS-RCSCL-86-225 Rev. 2 PAGE 3 OF 8 The proposed license amendment (Technical Specification change) addresses the action required when degradation has been detected in the roll expanded portion of steam generator tubes within the steam generator tubesheet.
Existing tube repair or plugging
- criteria, i.e.
current applications of USNRC Regulatory Guide 1.121, do not take into account the effect of the tubasheet on the external surface of the tube.
The presence of the tubesheet will enhance the integrity of degraded tubes in that region by precluding tube deformation beyond the expanded outside diameter.
Additionally, a portion of the roll expansion at the top and of the tube expansion is sufficient to preclude pullout of the tube during normal operation and postulated accident condition loadings if a tube were postulated to sever circumfer--
entially during plant operations in the portion of the tube covered by the proposed amendment.
Finally, the roll expansion of the tube into the tubesheet provides a barrier to significant leakage for through wall cracking of the tube in the expanded region.
The proposed change designates a portion of the tube for which tube degradation does not necessitate remedial action except as dictated for compliance with tube leakage limits as set forth in the J.
M.
Farley Nuclear Plant Technical Specifications.
As noted
- above, the area subject to this change is in the expanded portion of the tube within the tubasheet of the steam generators.
The length of expansion required to resist pullout for all postulated conditions, designated F*,
has been determined to be 1.54 inches.
Since the expansion of the tube above F*
is sufficient to preclude pullout of the tube, use of the F*
criteria does not depend on any determination of the condition of tube degradation in the portion of the tube below the F*
distance.
The addition of an addy current elevation location uncertainty allowance of 0.25 inch results in operational F* value of 1.79 inches.
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NS-RCSCL-86-225 Rev. 2 PAGE 4 OF 8 The possibility of tube repair by sleeving should not be considered a
reason to exclude use of the alternate tubasheet plugging
- criteria, but should be considered one of the options used to address degradation in the expanded region of the tube.
The disadvantages of tube plugs noted above also apply to some extent to sleeves.
Additionally, installation of sleeves involves some impact on addy current testing due to the changes in geometry at the ends, expansions of the sleeve, and the size of probe that can pass through the reduced diameter of the sleeve.
Note that the Technical Specifications for J. M. Farley Unit 2
do not currently contain provisions for repair of tubes
)
by sleeving.
The proposed amendment would modify Technical Specifications 3/4.4.6 Steam Generator Bases and 4.4.6 Steam Generator Surveil-lance Requirements which provide tube inspection requirements and acceptance criteria to determine the level of degradation for which the tube may remain in service.
The proposed amend-ment would add definitions required for the alternate plugging criteria and prescribe the portion of the tube subject to the criteria.
The proposed Technical Specification changes accompany this analysis.
The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging operations.
The proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam generator in LOCA analyses.
The proposed amendment would avoid loss of margin in reactor coolant system flow and therefore assist in assuring that minimum flow rates are maintained in excess of that required for operation at full power.
Reduction in the amount of tube plugging required can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.
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NS-RCSCL-86-225 Rev. 2 PAGE 5 OF 8 ANALYSIS l
l Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
1)
Operation of the Farley Nuclear Plant Unit 2 in accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an accident previously evaluated.
The supporting technical and safety evaluations of the subject criteria
[ Westinghouse WCAP 11306 Rev.
2, "Tubesheet Roll Region Plugging
- Criteria, J. M. Farley Nuclear Plant, Series 51 Steam Generators" (Proprietary),
WCAP 11314 Rev. 2, (Non Proprietary),
and SECL-86-381 Rev.2) demonstrate that the presence of the tubesheet will enhance the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter.
The resistance to both tube rupture and tube collapse is strengthened by the presence of the tubesheet in that region.
The result of the expansion of the tube into the tubesheet is an interference fit between the tube and the tubesheet.
Tube rupture can not occur because the contact between the tube and tubesheet does
(
not permit sufficient movement of tube material.
In a similar j
- manner, the tubasheet does not permit sufficient movement of tube material to permit buckling collapse of the tube during postulated LOCA loadings.
Additionally through analysis and testing, Westinghouse has demonstrated that the roll expansion above the F* distance is sufficient to preclude pullout of the tube from the tubesheet.
Even with the conservative assumption that a tube could completely sever circumferentially below the F*
- distance, test results demonstrate that pullout of the tube
)
NS-RCSCL-86-225 Rev. 2 PAGE 6 OF 8 is precluded under normal and postulated accident condition loadings.
This assumption is conservative as the PWSCC that has been observed in operating units has been typified as short and axially oriented.
A conservative allowance is added for eddy current elevation location uncertainty to determine the operational value of F*.
Relative to expected leakage, the length of roll expansion above F*
is sufficient to preclude significant leakage from tube degradation located below the F*
distance.
The existing Technical Specification leakage rate requirements and accident analysis assumptions remain unchanged in the unlikely event significant leakage from this region does occur.
As noted
- above, tube rupture and pullout is not expected for tubes using the alternate plugging criteria.
Any leakage out of the tube from within the tubesheet at any elevation in the tubesheet is fully bounded by the existing i
steam generator tube rupture analysis included in the Farley Nuclear Plant Final Safety Analysis Report.
The proposed alternate plugging criteria do not adversely impact any other previously evaluated design basis accident.
i 2)
The proposed license amendment does not create the possi -
bility of a
new or different kind of accident from any accident previously evaluated.
I Implementation of the proposed alternate tubesheet plugging criteria does not introduce any significant changes to the plant design basis.
Use of the criteria does not provide a mechanism to result in an accident outside of the region of the tubesheet expansion.
Any hypothetical accident as a result of any tube degradation in the expanded portion of i
the tube would be bounded by the existing tube rupture accident accident analysis.
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NS-RCSCL-86-225 Rev. 2 PAGE 7 OF 8 3)
The proposed license amendment does not involve a signifi-cant reduction in a margin of safety.
The use of the alternate tubasheet plugging criteria (F*)
has been demonstrated to maintain the integrity of the tube bundle commensurate with the requirements of Reg Guide 1.121 for indications in the free span of tubes and the primary to secondary pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation is any degradation in the tubesheet more than the F* distance below the bottom of the roll transition.
The safety factors used in the determination of the F* distance and the strength of degraded tubes are consistent with the safety factors in the
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ASME Boiler and Pressure Vessel Code used in steam generator design.
The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude significant leakage during normal and postulated accident conditions.
The allowance used for eddy current elevation location measurement uncertainty has been supported by previous experience and laboratory testing.
For axial or nearly axial indications in the tubesheet
- region, the tube and remains structurally intact further.
decreasing any potential for tube pullout.
For tubes with axial or nearly axial
- cracks, the strength of the tube relative to an axial load would not be reduced below the strength required to resist potenti51, axial loads.
In this
- case, leakage is the dominant consideration to determine the necessity of tube plugging or repairing.
Again, based on
- testing, using the alternate plugging criteria would not be expected to result in significant leakage from through wall cracks located below the F* distance.
/
NS-RCSCIr86-225 Rev. 2 PAGE 8 OF 8 Implementation of the alternate tubesheet plugging criterion will decrease the number of tubes which must be taken out of service with tube plugs or repaired with sleeves.
Both plugs and sleeves reduce the RCS flow margin, thus implemen-tation of the alternate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change does not result in a significant reduction in a
loss of margin with respect to plant safety as defined in the Final Safety Analysis Report or the bases of the plant technical specifications.
CONCLUSION
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Based on the preceding analysis it is concluded that operation of J.
M.
Farley Nuclear Plant Unit 2 in accordance with the proposed amendment does not result in the creation of an unreviewed safety question, an increase in the probability of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, nor reduce any margins to plant safety.
Therefore, j
the license amendment does not involve a Significant Hazards Consideration as defined in 10 CFR 50.92.
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ENCLOSURE 1 Tubesheet Region Plugging Criterion for the Alabama Power Coupany Farley Nuclear Station Unit 2 Steam Generators September 1986 Revised April 1987 l
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