ML20127B717
| ML20127B717 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/08/1993 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20127B714 | List: |
| References | |
| NUDOCS 9301130076 | |
| Download: ML20127B717 (18) | |
Text
4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 3.
A tube inspection (pursuant to Specification 4.4.5.4.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 inspection, this shall be recorded and an adjacent tube shall~
23,y p be selected and subjected to a tube-inspection.
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c.
The tubes selected as the second and third samplos:(if required by Table 4.4-2) during-each inservice inspection may be subjected to a-partial tube inspection provided:
l.
The' tubes selected for these samples. include the tubes from those areas of the tube sheet array where tubes with imperfec-tions were previously found.
2.-
The inspections include those. portions oflthe tubes.where-imperfections were previously_found.
The results of each sample inspection shall be classified into one 'of the following three categories:
Cateoory-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:l% 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 totalftubes; ins'pected are degraded.
tubes or more than 1% of the' inspected tubes are defective.
Note:
In all inspections, previously ~ degraded _ tubes;must exhibit significant (greater than 10%) further wall-penetrations to be included in the:above percentage calculations.
SEQUOYAH - UNIT'1 3/4 4-7 9301130076 930100 PDR ADOCK-05000327 p
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-SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6' Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent Inservice inspections shall be-performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under-AVT conditions, not including the preservice. inspection, result in all inspection results falling into the C-lLcategory or if two consecutive inspections demo'nstrate that previously observed degrada-tion 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 results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection-frequency shall be increased to at-least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval-may then be extended to a maximum of once per 40 months.
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.
l.
Primary-to-secondary tubes leaks (not' including leaks originating from tube-to-tube sheet welds)'in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating-Basis ~ Earthquake.
3.
A loss-of-coolant accident-. requiring actuation of the-engineered safeguards.
4.
A main steam line or. feedwater line break.
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SEQUOYAH - UNIT 1 3/4 4-8
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be con-sidered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occuring or either inside or outside of a tube.
3.
Degraded Tube means a tube containing imperfections greater than or equal to 20% of 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 exceeds the plugging limit. A tube containing a defect is defective.
LA Pluggihq Limit mea the impe ection dep at or bey d c.
which th tube shall e removed om servic and is equ yp g
o 40% of e nominal be wall th' kness.
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 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.
9.
Preservice Inspection means a tube inspection of the full length of each~ tube in each steam generator performed by eddy current techniques prior to service establish a baseline con-dition of the tubing.
This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
O kuu SEQUOYAH - UNIT 1 3/4 4-9 SEP 171980 l
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SURVEILLANCE REQUIREMENTS (Continued) b.
The steam generator shall be determined OPERABLE after completing the corres onding actg@4Mng-44mough-wa44-era limit %dl -te >etr-conta
_ tubes exceedin the plugging required by Table 4.4-2 ~
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4.4.5.5 Reports
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Followl g each i ervice \\irspection f steam nerator i bes, the number o tubes p gged in (e ch steam enerator hall be eported gg N d,7 he Commi gion with q 15 day b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to specification 6.9.2 within 12 months following completion of the inspection.
This Special Report shall include:
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, bb@h Results of steam generator tube inspections which fall into Category c.
C-3 shall be reported pursuant to Specification 6.6.1 prior to resump.
R40 tion 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.
d November 23, 1984 SEQUOYAH - UNIT 1 3/4 4-10 Amendment No. 36
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BASES i
i 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 the secondary coolant chemistry is not maintained within these limits, localized corrosion may i
likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary. coolant system and the secondary coolant system (primary-to-secondary-leakage : 500 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will hive an adequate margin of safety to withstand the. loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by' radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an
- i unscheduled inspection, during which the leaking tubes will be located and Iwseur plugged.
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'Inslage-type detects are unlikely with proper chemistry treatment of the
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secoridary coolant.
However, even if a defect should develop in service, it will be found during scheduled inservice-steam generator tube examinations.
Plugging will be required for_all t_gbes with -imperfections exceeding the plugging limithf J0X cM t-obOcnnal wcT' thickscaO Steam generator tube
' 'N' inspections of operatlng plants have ilemonstrated ihe capability to reliably.
- rauer detect degradation that has penetrated 20% of the original tube wall thickness.
G)
Whenever the results.of any steam generator tubing instrvIce inspection gggg f all into_ Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.6.1 prior to resumption of plant opera-R40 tion.
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 necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE
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l 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS
~E The RCS. leakage detection-systems required'by this specification are provided to monitor and detect. leakage-from the_ Reactor Coolant Pressure Boundary.
These detection systems are consistent with-the recommendations of-Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection ~-
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Systems," May-1973.
k November 13. 1989 SEQUUfAH - UH11 1 B 3/4 4-3 Amendment No._36 i
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REACTOR COOLANT SYSTEM SURVE!LLANCE REQUIREMENTS (Continued)
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1.
All nonplugged tubes that previously had detectable wall pene-trations (greater than 20%).
2.
Tubes in those a.
is where experience has indicated potential problems.
3.
A tube inspection (pursuant _to Specification 4.4,5.4.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
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inspection, this shall be recorded and_an adjacent tube _shall-2' sert be selected and subjected to a tube inspection.-
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4(e s, c.
The tubes 1 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 where1 imperfections were previously.found.
The results of each sample inspection shall be classified into-one of1the 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% and110% of the' total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are-degraded tubes or more than =l% of the inspected tubes are defectiv'e.
Note:
In all inspections,_previously degraded tubesimust exhibit:
significant (greater than 10%) further wall: penetrations to-be included in the above percentage ~ calculations <
SEQUOYAH - UNIT 2' 3/4 4-11 1
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SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.3 inspection Frecuencies - The above required inservice inspections of steam generator tubes snall be performed at the following-frequencies:
a.
The first inservice inspection shall be performed after 6 Effective-Full Power Months but within 24 calendar months of initial criti-cality.
Subsequent inservice. inspections shall be performed at intervals of.not less than 12 nor more than 24 calendar months after i
the previous inspection.
If two consecutive. inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results f alling into the C-1 category or -if two consecutive 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 resslts of the inservice insoection of a steam generator conducted 'in accordance with Table 4.4-2 't 40 month intervals f all in Category C-3. the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.
c.
Additional, unscheculea 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 conaitions:
1.
Primary-to-secondary tubes leaks (not including leaks originating from tubo-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A loss-of-coolant accident requiring actuation of the engineered safeguarJs.
4.
A main steam line or feedwater line break, k sedr N
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REACTOR COOLANT SYSTEM SURVEILt.ANCE REQUIREMENTS (Continued) 4.4.3.4 Acceptance Criteria a.
As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be con-sidered as imperfections.
2.
Deoradation 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 greater than or equal to 20% cf the nominal wall thickness caused by degradation.
4.
% Decradation means the percentage of the tube wall tnickness affected or removed by degradation.
ari imperfection of such severity that it exceeds 5.
Defect m a
the plug.
,9 limit.
A tube containing a defect is defective.
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Pluggin3N(imitmeans imperfect 1 depth at or eyond which's Ci
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11 be remove from service nd is equal 40%
t e tuce s b "' '
the nomina. tube wall th' kness.
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7.
Inserviceable describes the condition of a tube if it leaks or Iantains a defect large enough to affect its structural integ-rity 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.
q 3.
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.
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Luc SEQUOYAH - UNIT 2 3/4 4-13
SURVEILLANCE REQUIREMENTS (Continued) 9.
_Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques orior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
The steam generctor shall be determined OPERABLE af ter completing b.
the corresponding actions (pluq uall tub_ exceed gin the plugging limit $dfan-tffetaTRIEg-through-waWeracky) required by Table 4.4-2.
4.4.5.5
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\\numberb tubes plugged in each steam generator shqll be rep ~orted to the Commi tion withtq l5 days.\\
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b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification G.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
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l 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 Category c.
C-3 shall be reported pursuant to Specification 6.6.1 prior to resump-tion of plant operation. The written followup of this report shall R28 provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
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November 23, 1984 SEQUOYAH - UNIT 2 3/4 4-14 Amendment No. 2B
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BASES 3/4.4.5 STEAM GENERATORS j
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.
l The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, 8evision 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, manuf acturing 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 the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during WW operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons. per day per steam generator can readily be detected by radiation. monitors of steam generator l
blowdown.
Leakage in excess of this limit will require plant shutdown an_d an-unscheduled inspection, during which the leaking tubes will be located and mer plugged.
F unc Wastage-type defects are unlikely with proper chemistry treatment of the secondary-coolant.
However, even if a defect should develop in service, it-will be found during scheduled inservice steam generator tube examinations.
Plugging will bg mLquired_for all tubes _ with imperfections exceeding the plugging limit.(gi,_ic" of the tute nominal-waWthkkness) -Steam generator tube inspections of operatiWplants haVrdemonstrated the capability to:
reliably detect degradation that has penetrated 20% of the original tube wall ~
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thickness.
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Whenever the results of any steam generator tubing inservice inspection y,,,f.
fall into Category C-3, these results will be-promptly rep'orted to the Commission-pursuant to Specification 6.9.1 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 necessary.
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ENCLOSURE 2.
PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAll NUCLEAR PIANTS UNITS 1 AND 2 DOCKET NOS. 50-327 AND'50-328 l
("IVA-SQN-TS-92-10)
DESCRIPTION mfd JUSTIFICATION FOR REVISING SURVEILIANCE REQUIRENENTS l
4.4.5.2, 4.4.5.3, 4.4.5.4, 4.4.5.5 l
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Description _of_ Change '
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TVA proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2-technical specifications.(TSs) to incorporate new requirements associated 3
with steam generator (S/G) tube inspection and repair._ The new requirements establish alternate S/G tube plugging criteria (i.e., W* -
criteria) that take into account the reinforcing effect the S/G tubesheet 4
has on the external surface of expanded S/G tubes referred to as i
Westinghouse Explosive Tube Expansion (WEXTEX) expanded tubes. The-
. proposed changes are as follows:
- 1. - Add Surveillance Requirement (SR) 4.4.5.2.b.4 The SR provides a new requirement _for examining the full length of_the tubesheet region using bobbin coil for all_nonplugged, tubes with previously identified degradation.
The new SR requires examination of the W*
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region for all previously identified W*' tubes using a rotating pancake coil (RPC).
- i. Add SR 4.4.5.3.d The SR provides a new requirement for examining all S/Gs to which the W* criteria has been applied.
3.
Revise SR 4.4.5.4.a.6 The SR has been modified to change the current-definition of Elugging_ Limit to account for the W* criteria.
4.
Add SR 4.4.5.4.a.10 The SR provxdes a new' definition for the Ep.t. tom of..the WEXTEX' Transition (BWT).
5.
Add SR 4.4.5.4.a.11
. The SR Lprovides a new definition for. the:
M*_Diatance.
6.
Add SR 4.4.5.4.a.12 The'SR provides a'new definition for the W* lube.
7.
Revise SR 4.4.5.4.b The SR has been-modified'to'be consistent with the new definition'of the plugging limit.
- 8. : Revise SRL4.4.5.5.a-
-The-SR.provides a new reporting requirement _-
for W* tubes and clarifies the. tube plugging 1 L
report timing.:
l 9.
Add to Bases 3/4.4.5 The-proposed change adds information that' i
explains the basis for.the W* criteria'and_the.
l, limits ofiacceptability. The change.also:
deletes information that would no longer be consistent with the W* criteria.
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Renann_fsr__ Change TVA is proposing to change SRs 4.4.5.2 through 4.4.5.5 to reduce the need for repairing or plugging S/G tubes having indications that exceed the current TS depth-based plugging limit.
The alternate tube plugging criteria, or W* criteria, are based on maintaining structural and leakage integrity of tubes with indications in the WEXTEX region. The Westinghouse Electric Corporation has performed analyses to show that indications in the WEXTEX region meet Regulatory Guide (RG) 1.121 criteria for tube integrity and leakage in a faulted condition remain below allowable offsite radiation dose limits.
The proposed change would maximize the reactor coolant flow margin, and reduce the radiation exposure incurred in the process of plugging or repairing S/G tubes (approximately 0.060 manrem per tube of exposure would be saved for a plugging operation). Other benefits of not plugging tubesheet indications that meet the W* criteria would be a reduction in man-hours and potential impact to critical path time during refueling outages.
TVA's goal is to prolong the life of the S/Gs to the expected plant life of 40 years.
This goal is best achieved by proactive measures that defer or eliminate the need to replace S/Gs.
S/G replacement occurs because of the loss of tube plugging margin. TVA's goals of prolonged S/G life and reduction in personnel exposure while maintaining the SQN S/G plugging margin are the core reasons for this TS change request.
Jus.tilica_ tion _fot_Chaage SQN's current TS plugging limit of 40 percent through-wall applies throughout the tube length and does not take into account the reinforcing ef fect the tubesheet has on the external surf ace of the WEXTEX expanded tubes. The presence of the tubesheet acts to constrain the tube and complements tube integrity in that region by precluding tube deformation beyond the expanded outside diameter.
The resistance to tube rupture, tube collapse, and resistance to leakage is significantly enhanced by the tubesheet. RG 1.121. " Bases for Plugging Degraded PWR Steam Generator Tubes," (August 1976) provides a method acceptable to NRC staff for establishing safe limits of degradation for S/G tubing. The RG 1.121 criteria have formed the bases for SQN's current TS plugging limit. The RG conservatively considers only the f ree span tubing.
In the last outages, the SQN S/Gs have shown degradation within the tubesheet at and below the WEXTEX expansion transition region An evaluation was performed to assess leakage f rom cracks in the WEXTEX region during normal plant conditions and postulated accident conditions. To establish a conservative leak rate criteria for faulted conditions, the steam-line break (SLB) leakage, which has the most stringent radiological hazards, is applied using the higher feedwater line break pressure differential. The evaluation considers the leak rate to be a function of the differential pressure across the tube wall, the radial contact pressure between the tube and tubesheet, and the crevice length.
Prototypic WEXTEX tube-to-tubesheet joints and tubing with induced cracks were tested at various pressures and temperatures to i
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_3 provide data for a leak rate model with calculated leak rates for given crack lengths at various locations within the tubesheet.
These leak rate values are applied to W* tubes and the aggregate value limited per S/G dependent on the percent of tubes examined.
A total leak rate of I gallon per minute (gpm) per S/G remains consistent with the assumed S/G 1eak rate from SQN's SLB analysis (Final Safety Analysis Report, Table 15.5.4-1, Sheet 1).
The 1-gpm leak rate limit provides one of the TS criteria that determine if a tube is to be removed f rom service (i.e., TS plugging limit). The 1-gpm leak rate limit,-as it applies to the plugging limit, is applicable to the region within the W*
distance. Any indications below the W* distance are not expected to contribute more than 5 percent of the total SLB leak rate and are thereby neglected in the TS plugging-limit.
Testing was perfonned to determine the coef ficient of friction between the tube and the tubesheet. The axial forces on the tube that would cause movement of the tube during normal operation and during postulated accidents were determined. The radial forces on the tube during normal operation and during postulated accidents were applied to the coefficient of f riction and set equal to the axial forces to determine the necessary length of expanded tube required to preclude tube pullout from the tubesheet.
The following were considered in the determination of radial forcest a.
Tubesheet flex and the resulting dilation of the tubesheet hole.
b.
Thermal expansion of both the tubes and tubesheet.
c.
Primary and secondary. pressures of the S/G.
The W* criteria allow a degraded tube that meets a specific criteria to be lef t in service. A tube with a degraded region within the W* distance (the W* distance is the length of tubing.below the bottom of the WEXTEX transition (BWT] that must be demonstrated to be nondegraded) must have sufficient net cross-sectional area to transmit the axial load to' sound regions of the tube. No axial restraint is attributed to the degraded region; therefore, the axial loads are considered to be reacted by an.
aggregate W* distance of sound tube above and below degraded regions (degraded regions include crack length and' appropriate end effects).
Single axial cracks have no impact on the axial pull strength of the tube, but for circumferential-cracks, the-noneracked area must be-sufficiently large to carry the axial load.
Calculations have determined that a tube with a through-wall circumferential crack 197 degrees long and the balance of the tube wall 50 percent degraded would not separate under an applied axial load equal-to three times 1the normal operating conditions (the normal operating, condition is the limiting condition). Similarly, testing has shown.that-a tube with 5 axial cracks inclined 30 degrees or less from.the tube axis covering 185 degrees circumferential extent would not separate under an applied axial load three times the normal operating conditions.
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-s-The approach taken to develop the 9* criteria is to utilize the' general methodology of the staff-approved L* criteria for hardroll expansions and adapt the methods for WEXTEX expansions. The criteria utilize a length below the top of the tubesheet (W* distance) of nondegraded tubing to resist pull-out forces and to-limit leakage. When degradations are found, the distance is increased-such that an aggregate W* distance of nondegraded tubing is ensured to resist-the pull-out forces.. Since WEXTEX expansions have lower tube-to-tubesheet contact forces than hardroll expansions, (L*) limited leakage is possible under the SLB conditions and a requirement is made to calculate and limit the total SLB leakage for indications lef t in service.
Extensive operating experience in Europe.with free span axial primary water stress corrosion cracking cracks has demonstrated negligible leakage during normal operating conditions.
TVA's proposed TS change is consistent with RG 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," and requires all tubes in all S/Gs to which the W* criteria have been applied.
to be inspected using the RPC probes at all future in-service inspections. This requirement ensures that at the beginning of each fuel
+
cycle, S/G tube indications that are to remain in service still meet the W* criteria and do not reduce the factor of safety.
Based on analyses and the results of extensive testing, TVA concludes that plant operations in accordance with the-proposed changes are justified.
Environmental 1mpac.t Evaluntion The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would nott-1.
Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's. testimony to the~ Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and. Licensing Board.-
2.
Result in a significant change in effluents or power levels.
3, Result in matters not previously reviewed in the licensing basis:for-
- SQN that may.have a significant environmental impact.
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1 ENCLOSURE 3
' i PROPOSED TECHNICAL SPECIFICATION CilANGE SEQUOYAll NUCLEAR PLANTS-UN1TS -1 AND 2-1 DOCKET NOS. 50-327 AND 50-328 (IVA-SQN-TS-92-10)
DETERMINATION OF NO SIGNIFICANT llAZARDS CONSIDERATIONS 9
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Significant 11azards Conside=2*4cn Analysis TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c).
Operation of SQN in accordance with the proposed amendment will nott (1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
TVA proposes to modify the SQN Unit 1 and Unit 2 TSs to incorporate alternate steam generator (S/G) tube plugging criteria (i.e., W*
criteria). The W* criteria take into account the reinforcing offect the tubesheet has on the external surface of an expanded S/G tube.
TVA is requesting this change tot (1) maximize reactor coolant system flow, (2) maximize tube plugging margin to prolong S/G life,
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and (3) reduce occupational radiation exposure, man '.ours, and S/G l
critical path time.
Tube-bundle integrity will not be adversely affected by the' implementation of the W* tube plugging criterza S/G tube burst or collapse cannot occur within the confines of the tubesheet; therefore, the tube burst and collapse criteria of Regulatory Guide (RG) 1.121 are inherently met. The W* distance is shown by analyses i
and test results to retain the tube within the tubesheet during all-plant conditions, thereby precluding an event with consequences similar to a postulated tube rupture event.
For S/G tubes with indications that are returned to service as a result of the application of the W* criteria, postulated steam-line break (SLB) i leak rates are required to remain within the Final Safety Analysis Report (FSAR) limits.
e The W* distance has been shown to-restrict-primary-to-secondary leakage below current TS limits during all plant. conditions.
In the unlikely event that a through-wall crack develops and results in primary to secondary-leakage, the leakage characteristics and trends would allow safe shutdown of the plant.- Relative to the potential for SLB leakage, the distribution of cracking within the W* distance
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will be limited such that expected leakage and. associated radiological consequences will not exceed the FSAR limits.
In conclusion. the incorporation of the W* criteria into SQN_TSs
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maintains existing design limits and does not involve a significant
- increase in the probability or consequences ofian-accident previously evaluated.
(2) Create the possibility of a new or.different kind of accident from any previously analyzed.
Tube-bundle integrity is expected to be maintained during all plant conditions upon implementation of the proposed alternate tubesheet plugging criteria. Use of-the criteria does not induce a new mechanism that would-result in a different kind of accident from those previously analyzed.
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Even with the limiting circumstance of a complete circumferential-separation of a tube occurring below the W* distance, S/G tube pullout is precluded and leakage is predicted to be maintained within the FSAR limits during all plant conditions. Administrative leak rate limits and leakage trending are in place at SQN to address the potential for plant operation with circumferential cracking.
(3)
Involve a significant reduction in a margin of safety.
Upon implementation of the W* criteria, operation with potential cracking in the WEXTEX expansion region of the S/G tubing is predicted to meet the criteria of RG 1.121 and RG 1.83 and thereby-meet the-requirements of General Design Criteria 14, 15,-.31 and 32.
Calculations have shown that tubes.within the W* distance with circumferential cracks having arc lengths less than 92 degrees (beginning of the cycle) would not pull out during a postulated SLB event. The most significant_effect would be a possible Increase in-leakage following an SLB event.
Excessive leakage during an SLB event.is precluded since-calculations and-testing have shown that-the potential distribution of cracks both within and below the W*
7 distance would not exceed Icakage limits during all plant conditions.
Consequently, the margin of safety would not be significantly reduced.
Neglected in the above evaluation is the leakage restriction and tube restraint offered by corrosion product buildup in the tube-to-tubesheet crevice. This buildup will restrict leakage to below the currently. calculated amounts, in addition ~to adding to the tube pullout resistance, as shown by test results.
The addition of a new TS requirement for inspecting tubes using a rotating pancake coil probe will also contribute to the identification of any degradation mechanism that is not readily detectable using bobbin probe.
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