ML20138G958
| ML20138G958 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/26/1996 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20138G950 | List: |
| References | |
| GL-95-05, GL-95-5, NUDOCS 9701030077 | |
| Download: ML20138G958 (37) | |
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1 Unit 1 Technical Specification Page Markips i
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i 9701030077 961226 PDR ADOCK 05000348 P
BEACTom C00LAk7 SY3fD(
sutvtILLANCE REQUIRD4Dr78 (Cantanued) a, 1.
All nenpluggsd tubes that prsvacusly had catectable well penetrations greater than 204.
i 2.
Tubes in those areas where experience has indicated potential problems.
3.
At least 34 et the total nupher of sleeved tubes an all three steam generators er 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 4.4.4.4.a.01 shall be perferined on each selected tube.
If any selected tube does not permit the passage of j
the oddy 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.
i 2,L adications left in servies as a w.'sult.f apcicati.n.f th. tub. su,, ort, late g
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5.
_m I gg r
r
,1 ;; t
--1 ____ shall bjeaspected by bobbia ceti I
j
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. refueling outages.
probe during 6,H fvfJ/6, The tubes selected as the seceed and third sasyles (if c.
seguired my Table 4.4-21 during each inservice inspecties may be subjected to e partial tube inspection provided:
4 J
1.
The tubes selected for these samples include-the. tubes.
l from these areas of the tube sheet array where tubes with taperfections were previously found.
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2.
The inspections include these portions of the tubes i
where imperfections were previously found.
{
tatie the steam gengator tube / tube support plate,i _;,' gncriteria requireE100 percent bobbia coilj l
/
d.
imammeties for hot-leg i - ;; - ;'
" _ - __ _ __-_' -_ and glggp cold-le(*1stersections down to the lowest cold-leg tube e*f f
1 support plate with known outside diameter stress corresten
- g indications. The deternanaries ofPtube T
cracking (ODSCC) support plate intersections having 000CC indications shall 1
I i
he based en the performance et at. least a 20 percent randes sampling of tubes inspected ever their. full length.
The results of each sample inspection shall be classified into one of i
the fellowing three categories:
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(calciIy 3
rARI.ET-UNIT 1 3/4 4-10 AMDfDMDrT NO.M,11 1
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atACTom cooLA#rT SYSToi
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suavt!LLANCE hE0uthosprTs (Centanuedi e,
4.4.6.4 Acc*ptanee Criterte a.
As used in this Specification:
1-
!aperfection osans an onception to the dimenstens, f an&sh er contour of a tube er sleeve from that i
I required by f abrscation drawings er specttteetions.
Eddy-current testang indications below 20% of the nominal well thackness, it detectable, may be censidered as impeafoctions.
l 2.
Deeredation means a service-induced cracking, wastage, wear er geceral certesten occurring en either inside or outside of a tube or sleeve.
i i
3.
Deeraded Tube means a tube, including the sleeve if the tube has been repaired, that contains i
imperfections greater than er equal to 20% of the i
nominal wall thickness caused by degradation.
4.
4 Deeradottee means the pereestage of the tube er sleeve well thickness affected or removed by degradaties.
S.
means an imperfection of such severity that it ese the plugging er repair limit. A tube or sleeve containing a defect is defective.
6.
Pluestas or hesair Limit means the taperfection depth at er beyond which the tube shall be repaired (i.e.,
sleeved) er removed from service by plugging and is greater than er equal to 40% of the muninal tube well thickness. For a tube that has been sleeved with a mechanical jotat sleeve, through us11 penetration of greater than er equal to all et sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging. For a tube that has been sleeved with a welded joint sleeve, through well penetraties greater than er equal to 316 et sleeve seminal well thickness in the sleeve between the weld joints requires the tube to be removed f rom service by
,A plugging. This defiattien does met apply to tube summert plate intersections for _ which the voltage-L I
I hasee, _ _. _ __."nitaria are betag applied, mefor to 4.4.6.4.a.11 for the MT.Minit applicable to the.e intersecusas.j 7.
Unserviceable describes the condities of a tube er sleeve if it leaks er contains a defect large enough to affect its structural integrity in the event of an operettag Basis Earthquake, a less-et-seelaat accident, or a steam line er foodwater line break as specified in 4.4.6.3.c, above.
FAALEY-UNIT 1 3/4 4-12 AMDf DMDrf NO. 0 U O. ".
Mv444, 117
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SURYt!LLANCE REQUIREMDtTS ICcntanutdt
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O.
Tube Inspectnen means an anspection of the steam ganarator tuna f ree the paant of entry thst leg siset i
cespletely around the U-bend to tha tcP suppstt of ti.
i cold leg.
For a tube that has been repatred by sleeving, the tube inspection should anclude the sleeved portion of the tube.
9.
Tube Repair refers to eschanical sleeving, as described by Westinghouse report WCAF-11110, Rev. 1, or laser welded sleevang, as described by Westinghouse i
. report WCAp-12612, whica is used to maintain a tube in service er retura a tube to service. This includes j
the removal of plugs that were installed as a j
corrective er preventive measure.
1 i
10.
Preserviec Inspecties esans an inspection of the full length of each tube in each steam generator performed by oddy ci trent techniques prior to service to establish a baseline condition of the tubing. This inspectisen shall be perferned af ter the field hydrostatic tw and prior to initial PowtA OPERAT!cel usingt the equipmaat and techniques espected to be used during subsequent inservice inspections.
a
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11.
Tube Sun ort el.co r:-- :-Ms-it is us.d f.c the u/w 600 I
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I dianesitten eMteam generater tube for continued service taas an esimeriencineanutside diameter stress eorresten cracking eenrines within the thickness of s
fQ the, tube suPrert, plates.,]_______..._.y f
/M.sA,d ow 4
DM At tub
- aupF*Et plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described belows i
q a.j[ tress corr $ attributed to outside diameter gradation f
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s sies cracking within the bounds of d{g the tube support 21ste wig) bobbin voltage less than er equallag.0 voltQp11 he a11ewed to remain sa servres.
b.
A gradattee ttributed to outside diameter I
h ftp pth
. /,
stress corrosies cracking within the bounds of
!/4115 the tube supaget ple with a bobbia voltage greater thug.O volt ill be repaired er av,,..
except as not la 4. 4. 6. 4.a.11.c below.
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gndications of potential adation attributed to outside diameter stress resten cracking within the bounds of the t PPert g3 ate with n k o / P / W . N a hobbia vele
.erester p 0 volte but less fj than er equal ite..:
2-mar eemain f a service g
a rotattag ----t rii inspectica does not Wma i
/ndicatians of outside amo er stress corrosion cracking Mradation 0
with a bobbia valtsgo greater than
- -id 11 he plugged er repaired.
rAmt.r?-Uwrv 1 3/4 4-12a AMD8ts<DfT No.*Med.117 g
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I SURVEILIANCE REQUIREMDtTS (C:ntinusdi I
l b.
Thz stcan generster shall bs determaned CPtRASLt af ter completing ths cocroopsnding actacne splug er repaar et all i
tubes exceeding the pluggang er repair 11 mitt reguared by
,i Table 4.4-2.
4.4.6.S Reports I
a.
Following each inservice anspection of steam generator tubes, the number of tubes plugged or repaired in each steam generater shall be reported to the Ceaunission within 15 days j
of the completten of the plugging er repair ef fort.
d i
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 aonths following the completion of the inspection. This Special Report shall include:
4 1.
Number and eatent of tubes and sleeves inspected.
2.
Location and percent of well-thickness penetration for i
]
each indication of an igerfecties.
l 3.
Identification of tubes plugged or repaired.
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c.
Results of steam generater tube inspections which fall into i
Category C-3 shall be eensidered a REPORTABLE EYDrf and j
shall be reported pursuant to 10CFRSS.73 prior to resumption of plant operation. The writter. report shall provide a description of investigations conducted to determine the cause of the tube degradaties and corrective sessures taken 3
to prevent recurrence, i
d.
For iglementation of the weltage-based repair criteria to i
tube support plate intersections, notify the NRC staff prior to returning the steam generators to service (Mode il should any of the following ceeditions arise:
p 1.
b estimated leakage based on the actual end-of-cyc1 voltage distributies would have escoeded the leak 7
limit ifor the postulated nata steam Hoe brook utilising licensing basis assugtions) ductag the Lyrevious operating cycle.
4 2.
It circumferential crack-like indications are detected at the tube support plate intersections.
3.
If indications are identified that eatend beyond the ceafines of the tube support plate.
p b the calculated.s,conditiemal burst probability
/feW {
4.
escoeds 1.0 a 10 motify the NRC and provide an assessment of the safety significance of the occurrence.
2k FARLEY-Ult!T 1 3/4 4-13 AMD8DMENT #0.Kr44,117 i
v 3/4.4.6 STIAM GtytRAT*Rs Tha Surveillance R:quirements for inspsetica of the etsaa g:norater tubas ensure that the struttural integrity of this portion of ths RCS will be maintained. The program for inservice inspectaen of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revisten 1.
Inservice inspection of steam generater 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 thM 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 correct &ve measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within these chemistry limits found to result in negligible corresten of the steam generater tubes. !! the secondary coolant chemistry is not maintained within these limits, localised corrosien may likely result in stress corrosion cracking. The estent of cracking during plant operation would be limited by the limitation of steam generater tube leakage between the primary seelant system and the secondary coolant system tyrimary-to-secondary leakage = 140 gallons per day per steam generatori.
Cracks having a primary-to-secondary leakage less than this limit during operaties will have as adequate margia of safety to withstand the needs imposed during nesmal operaties and by postulsted accidents. Operational leakage of this magnitude saa be readily detected by asisting Farley Unit 1 radiatica maaitors. Imakage in excess of this limit will require plant shutdown and an unscheduled bepecties, during which the leaking tubes will be located and pluggsd er repaimd.
IThe repair limit for 00 SCC at tube support plate intersections is based on the h
analysis contained in NCAF-12071, Revision 2.
"J. N. Farley Units 1 and 2 SG Tube Flugging criteria for 00 SCC at Tube Support Plates,' and documentation contained in EPRI Report TR-100407, Revision 1, "PWR Steam Generator tube Repair Limits - Technical Support Document for Outside Diameter Stress Cerrosion cracking at Tube Support Plates." The application of this criteria is based on limiting primary-to-secondary leakage during a steam line break to '
naure the applicable Part 100 limits are not onceeded.
J Wastage-type defects are unlikely with proper chemistry treatment of the secondary coelaat. Bewever, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube esaminations.
Flugging er repair will be required for all tubes with imperfections onceeding 404 et the tube nominal wall thickness. If a sleeved tube is found to have through wall penetration of greater than er equal to all for the mechanical sleeve and 376 for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 314 and 374 timits are derived f rom R.G.
1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of well degradaties met be evaluated can be susumarised as follows:
Ik2M F"AR1.EY-UNIT 1 3 3/4 4-3 AMD&DDENT 100
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i REACTOR C00LAN* SYSTEM l
BASES 1
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a.
Mechanical 1.
Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve pluggang 11 mat.
2.
Indication of tube degradation of any type including a l
complete guillotine break in the tube between the bottom of th,e upper joint and the top of the lower roll expansion does i
not require that the tube be removed f rom servtce.
- 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 lumat applies to these creas also.
4.
The tube plugging limit continues to apply to that porttoo of the tube above the top of the upper joint.
b.
Laser Welded 1.
Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.
3.
At the weld joint, cegradation must be evaluated in both the sleeve and tube.
4.
In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the 1 Lait of the sleeve inspection.
5.
The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.
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 well 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 10 CFR 50.13 prior to resumption of plant operation. Such cases will be considered by the Cammission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision to the Technical specifications, if nece s s a ry.
FARLEY-UNIT 1 53/4 4-3 AND;DMENT NO.
l 3 3; 72,85,106
t Insert A s
d.
If an unscheduled mid-cycle inspection is performed, l
the following mid-cycle repair limits apply _instead of the limits identified in 4. 4. 6. 4.a.11.a, L
4.4.6.4.a.11.b, and 4.4.6.4.a.11.c.
.The mid-cycle l
repair limits are determined from the following i
equations:
i l-Vst Vmu= 1.0 + NDE + Gr [ CL-At )
CL i
Van =Vmu-[Vun-V e), [ CL-At ]
t CL i
where:
I Vuu upper voltage repair limit
=
lower voltage repair limit j
Von
=
mid-cycle upper voltage repair limit based Vme
=
on time into cycle mid-cycle lower voltage repair limit based Van
=
on V u and time into cycle length of time since last scheduled At
=
inspection during which Vug and V g were t
implemented cycle length (the time between two i
CL
=
scheduled steam generator inspections) structural limit voltage l
Vst
=
average. growth rate per cycle length i
Gr
=
95-percent cumulative probability i
=
allowance for nondestructive examination uncertainty (i.e.,
a value of 20-percent l
has been approved by NRC)
Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.6.4.a.11.a, 4.4.6.4.a.11.b, and 4.4.6.4.a.ll.c.
t i
t i
i w
l l
_ _... -. _ -. - _ _. ~., _
m l.
Inscrt D i
f 1.
If estimated leakage based on the projected end-of-cycle (or i
if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from l
the licensing basis dose calculation for the postulated main
{
steam line break) for the next operating cycle.
l l'
I
. Insert c j
L 4.
It indications are identified'at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
j i
)
5.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the l
E actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the f
safety significance of the occurrence.
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Inscrt D The voltage-based repair limits of 4.4.6 implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generatorr (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections. The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG.
Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of 4.4.6 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent i
derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650 *F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit, Ven, is determined from the structural voltage limit by applying the following equation:
Vun = Vn - Va - Vee I
where Vu represents the allowance for flaw growth between inspections and Voc
)
represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit is contained in GL 95-05.
The mid-cycle equation in 4.4.6.4.a.11.d should only be used during unplanned
)
inspections in which eddy current data is acquired for indications at the tube I
support plates.
4.4.6.5 inplements several reporting requirements recommended by GL 95-05 for situations in which the NRC wants to be notified prior to returning the SGs to i
service.
For the purposes of this reporting requirement, leakage and l
conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for j
the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, j
then the results of the projected EOC voltage distribution should be provided per the GL section 6.b(c) criteria.
4, p
i Unit i Technical Specification Pages Replacement Pages Page 3/4 4 10 Replace Page 3/4 4-12 Replace Page 3/4 4-12a Replace Page 3/4 4-12b insert Page 3/4 4-13 Replace Page B 3/4 4-3 Replace Page B 3/4 4-3a Replace Page B 3/4 4 3b insert I
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REACTOR COOLA" T SYSTEM SURVEILLAN^
REQUIREMENTS (Continued) l.
All nonplugged tubes that previously had detectable j
wall penetrations greater than 20%.
2.
Tubes in those areas where experience has indicated potential problems.
3.
At least 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, l
whichever is less.
These inspections will include both the tube and the sleeve.
I 4.
A tube inspection (pursuant to Specification 3
4.4.6.4.a.8) shall be performed on each selected tube.
J 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.
l S.
Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.
j c.
The tubes selected as the second and third samples (if j
required by Table 4.4-2) during each inservice inspection
]
i may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes i
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.
d.
Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil l
inspection for hot-leg and cold-leg tube support plate i
intersections down to the lowest cold-leg tube support plate l with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.
I The results of each sample inspection shall be classified into one of the following three categories:
FARLEY-UNIT 1 3/4 4-10 AMENDMENT NO.
i
m. _
a 6
+
SURVEILLANCE REQUIREMENTS'(Continued) 4.4.6.4 Acceptance Criteria t
a.
As.used in this Specification:
l 1.
Imperfection means an exception to the dimensions,
[
l finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications Lelow 20% of the i
nominal wall thickness, if detectable, may be considered as imperfections.
L 2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside i'
or outside of a tube or sleeve.
3.
Degraded Tube means a tube, including the sleeve if f
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.
i S.
Defect means an imperfection of such severity that it exceeds the plugging or. repair limit. A tube or sleeve containing a defect is defective.
i 6.
Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e.,
l sleeved) or removed from service by plugging and is l
greater than or equal to 40% of the nominal tube wall thickness.
For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater than or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging.
For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 37% of sleeve nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging.
This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are beis.g applied.
Refer to i
4.4.6.4.a.11 for the repair limit applicable to these intersections.
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.
i FARLEY-UNIT 1 3/4 4-12 AMENDMENT NO.
[
^
REACTOR COOLANT SYSTEM' SURVEILLANCE REQUIREMENTS (Continued)
)
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.
l l
9.
Tube Repair refers to mechanical sleeving, as i
described by Westinghouse report WCAP-lll?8, Rev. 1, j
or laser welded sleeving, as described by Westinghouse report WCAP-12672, 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.
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.
Tube Support Plate Repair Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly
)
axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube i
l support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:
a.
Steam generator tubes, whose degradation is
)
attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage less than or equal to the lower voltage repair limit [2.0 volts), will be allowed to remain in service.
1 b.
Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [2.0 volts), will be repaired or plugged except as noted in 4.4.6.4.a.11.c below.
EARLEY-UNIT 1 3/4 4-12a AMENDMENT NO.
l i
REACTOR COOLANT SYSTEM SURVE1LLANCE REQUIREMENTS (Continuad) c.
Steam generator tubes, with indications of pote,tial degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [2.0 volts), but less than or equal to the upper voltage repair lindt*, may remain in service if a rotating probe inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage I
gretter than the upper voltage repair lindt*,
will be plugged or repaired.
i d.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair.'mits apply instead of the limits identified in 4.4.6.4.a.11.a, 4.4.6.4.a.11.b, and 4.4.6.4.a.11.c.
The mid-cycle repair lindts are determined from the following equations i
Vst j
Vme= 1.0 + NDE + Gr [ CL-At )
CL
[
t Ven=Vmg-[Vun-V n] [ CL-At ]
l t
CL where:
upper voltage repair limit i
Vpa
=
t lower voltage repair limit Vu
=
mid-cycle upper voltage repair limit i
Vmg
=
based on time into cycle mid-cycle lower voltage repair lindt Van
=
based on Vma and time into cycle i
At length of time since last scheduled
=
inspection during which Vuu and Vp; j
were implemented cycle length (the time between two CL
=
i scheduled steam generator
)
inspections) structural limit voltage i
V,o
=
average growth rate per cycle length Gr
=
95-percent cumulative probability NDE
=
allowance for nondestructive examination uncertainty (i.e.,
a l
value of 20-percent has been l
approved by NRC)
(
Implementation of these mid-cycle repair limits should fo3)ow the same approach as in TS 4.4.6.4.a.11.a, 4.4.6.4.a.11.b, and 4.4.6.4.a.11.c.
The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
1 FARLEY-UNIT 1 3/4 4-12b AMENDMENT NO.
i I
i I
1
+
1 e.
SURVEILLANCE REQUIREMENTS (Continusd) b.
The steam generator shall be determined OPERABLE after i
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 i
a.
Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam i
1 generator shall be reported to the Commission within 15 days of the completion of the plugging or repair effort.
l 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 j
months following the completion of the inspection.
This Special Report shall include:
4 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.
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. The written report shall provide a i
description of investigations conducted to determine the cause of the tube degradation and corrective measures taken
[
to prevent recurrence, d.
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior L
to returning the steam generators to service (Mode 4) should i
any of the following conditions arise:
1.
If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.
l 2.
If circumferential crack-like indications are detected at the tube support plate intersections.
3.
If indications are identified that extend beyond the confines of the tube support plate.
4.
If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
l S.
If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage i
distribution exceeds 1 x 10", notify the NRC and I
provide an assessment of the safety significance of the occurrence.
FARLEY-UNIT 1 3/4 4-13 AMENDMENT NO.
7
.._ _. ~.
.c
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- +
BASES i
l 3/4.4.6 STEAM GENERATORS 6
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 the secondary coolant chemistry is not maintained within these lindts, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the lindtation of steam generator tube leakage between the primary coolant system and the secondary coolant system l
(primary-to-secondary leakage = 140 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during I
operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operational l
leakage of this magnitude can be readily detected by existing Farley Unit 1 radiation monitors.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.
i The voltage-based repair limits of 4.4.6 implement the guidance in GL 95-05 l
and are applicable only to Westinghouse-designed steam generators (SGs) with j
outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube l
support plate intersections.
The voltage-based repair limits are not
(
applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG.
Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the l
thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of 4.4.6 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
i FARLEY-UNIT 1 B 3/4 4-3 AMENDMENT NO.
I r
i l
, _ ~ -
?
e.
l.
BASES The voltage structural lindt is the voltage from the burst pressure / bobbin voltage correlation at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance for tubing material properties at 650 'F (i. e., the 95-percent LTL curve). The voltage structural limit must i
be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit, l
Vun, is determined from the structural voltage limit by applying the following equation:
Vug = Vn - Vc,- V,x i
where Va, represents the allowance for flaw growth between inspections and Vnx represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit is contained in GL 95-05.
The mid-cycle equation in 4.4.6.4.a.11.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
4.4.6.5 implements several reporting requirements recommended by GL 95-05 for situations in which the NRC wants to be notified prior to returning the SGs to service.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for j
the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b(c) criteria.
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 or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness.
If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G.
l 1.121 calculations with 20% added for conservatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can i
be summarized as follows:
FARLEY-UNIT 1 B 3/4 4-3a AMENDMENT NO.
1 i
l i
BASES i
I
- a. Mechanical I
1.
Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.
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 lindt applies to these areas also.
i 4.
The tube plugging limit continues to apply to that portion of the tube above the top of the upper joint.
- b. Laser Welded 1.
Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.
3.
At the weld joint, degradation must be evaluated in both the sleeve and tube.
4.
In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.
5.
The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint, 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 10 CFR 50.73 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 i
requirement for analysis, laboratory examinations, tests, additicnal eddy-i current inspection, and revision to the Technical Specifications, if necessary.
t EARLEY-UNIT 1 B 3/4 4-3b AMENDMENT NO.
l 1
i
i Significant Hazards Evaluation Voltage-Based Repair Criteria
o i
Joseph M. Farley Nuclear Plant - Unit 1 i
Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking Sienificant Hazards Consideration Analysig
)
DESCRIPTION OF CHANGES As required by 10 CFR 50.91(a)(1), an analysis is provided to demonstrate that the proposed license amendment to implement the voltage-based repair criteria for tube support plate elevations in accordance with Generic Letter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Jorrosion Cracking," involves no significant hazards. The voltage-based repair criteria involve a correlation between eddy current bobbin probe signal amplitudes (voltage) and the tube burst and leakage capabilities.
Specifically, crack indications with bobbin probe voltages less than or equal to 2.0 volts, regardless of indicated depth, do not require remedial action if postulated steam line break leakage can be shown to be acceptable. Inspections will also be performed to ensure other forms of degradation are not occurring at the tube support plates and that cracks are not being masked at tube support plates by other factors.
The proposed amendment would modify Technical Specification 3/4.4.6 " Steam Generators" and its associated bases. He steam generator repair limit will be modified to clarify that the appropriate method
' for determining seniceability for tubes with outside diameter stress corrosion cracking at the tube support plate is by a methodology that more reliably assesses structural integrity. For Unit I, the operational leakage requirement has previously been modified to reduce the total allowabic primary-to-secondary leakage for any one steam generator from 500 gallons per day to 140 gallons per day. In addition, the technical specification limit for specific activity of dose equivalent 1"' and its transient dose equivalent I"'
reactor coolant specific activity has previously been reduced by a factor of 2 in order to increase the allowable leakage in the event of a steam line break.
EVALUATION Steam Generator Tube Intearity I
In the development of the voltage-based repair criteria, R.G.1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes,' and R.G.1.83, "Inser ice Inspection of PWR Steam Generator Tubes," were used as the bases for determining that steam generator tube integrity considerations are maintained within acceptable limits. R.G.1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria (GDC) 2,14,15,31, and 32 by reducing the probability and consequences of steam generator tube rupture through determimng the limiting safe conditions of tube wall degradation beyond which tubes with unacceptable cracking, as established by insenice inspection, should be removed from senice by plugging or repair. This regulatory guide uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME Code. For the tube support plate elevation degradation occurring in the Farley steam generators, tube burst enteria are inherently satisfied during normal operating i
conditions by the presence of the tube support plate. He presence of the tube support plate enhances the l
integrity of the degraded tubes in that region by precluding tube deformation beyond the diameter of the i
l
- Significant Hazards Evaluation Page 2 Voltage-Based Repair Criteria l
drilled hole. Analyses in WCAP-12871 show that for open crevices with as designed gaps, the tube l
l suppor: plate may not function to provide a similar constraining effect during accident condition loadings.
He WCAP-12871 analyses for Farley Unit I with corroded and packed crevices, as confirmed by bobbin coil inspection, show that the tube support plates would not be significantly displaced even under steam line break loading conditions. For conservatism in the voltage-based repair criteria, no credit is taken in the development of the repair criteria for the presence of the tube support plate during accident condition loadings. Conservatively, based on the existing data base, burst testing shows that the safety requirements for tube burst margins during accident condition loadings can be satisfied with bobbin coil signal amplitudes several times larger than the proposed 2.0 volt voltage-based repair criteria, regardless of the depth of tube wall penetration of the cracking. R.G.1.83 describes a method acceptable to the NRC staff for implementing GDC 14,15, 31, and 32 through periodic inservice inspection for the detection of significant tube wall degradation.
Upon implementation of the voltage-based repair criteria in accordance with Generic Letter 95-05, tube leakage considerations must also be addressed. It must be determined that the cracks will not leak excessively during all plant conditions. For the voltage-based repair criteria developed for the steam generator tubes, no leaxage is expected during normal operating conditions even with the presence of through-wall cracks. This is the case as the stress corrosion cracking occurring in the tubes at the support plate elevations in tne Farley steam generators consists of short, tight, axially oriented micro cracks often separated by ligaments of material. No leakage during normal operating conditions has been observed in the field for crack indications with signal amplitudes less than 7.7 volts in a 3/4 inch tube. Voltage correlation to 7/8 inch tubing size would result in an expected voltage of about 10 volts. Relative to the expected leakage during accident condition loadings, the limiting event with respect to primary-to-secondary leakage is a postulated steam line break event. For 7/8 inch tubing, the data supports no leakage up to 2.8 volts and a lo Vobability ofleakage between 2.8 and 6.0 volts. He threshold of significant 4
leakage (2 0.3 liter / hour or 10 gpm) in a 7/8 inch tube diameter is about 6 volts.
Mditional Considerations He proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging or repair operations. He proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam generator by keeping structurally sound tubes in service and not unnecessarily plugging or sleeving them and, therefore, assist in demonstrating that muumum flow rates are maintained in excess of that required for operation at full power. Reduction in the amount of tube plugging and sleeving can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.
ANALYSIS In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendment is 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.
Conformance of the proposed amendment 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:
i 4
- Significant Hazards Evaluation Page 3 Voltage-Based Repair Criteria 1)
Operation of Farley units in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Testing of model boiler specirnens for free standmg tubes at room temperature conditions shows j
burst pressures as high as approximately 5000 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 26.5 volts. Burst testing performed on pulled 2
tubes, including tubes pulled from Farley Unit 1, with up to 7.5 volt indications show burst pressures in excess of 5300 psi at room temperature. As stated earlier, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate.
Furthermore, correcting for the effects of temperature on material properties and nummum strength 4
levels (as the burst testmg was done at room temperature), tube burst capability significantly exceeds the R.G.1.121 criterion requiring the maintenance of a margin of 1.43 times the steam line break pressure differential on tube burst if through-wall cracks are present without regard to the presence of the tube suppoit plate. Considering the existing data base, this criterion is satisfied with bobbin coil indications with signal ampm.es over twice the 2.0 volt voltage-based repair criteria, regardless of the indicated depth measurement. His structural limit is based on a lower 95% confidence level limit of the data at operating temperatures. He 2.0 volt criterion provides an extremely conservative margin of safety to the structural limit considering expected growth rates of outside diameter stress corrosion cracking at Farley. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation.
However, relative to expected leakage during normal operating conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as compared to the 2.0 volt proposed voltage-based tube repair limit). Hus, the proposed amendment does not involve a significant increase in the probability or consequences of an accident.
Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the emironment are Loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses.
De offsite dose analyses for the other events which model primary-to-secondary leakage and steam releases from the secondary side to the emironment assume that the secondary side remains intact.
De steam generator tubes are not subjected to a sustamed increase in differential pressure, as is the case following a steam line break event. His increase in differential pressure is responsible for the postulated increase in leakage and associated offsite doses following a steam line break event.
~
In additim the steam line break i. vent results in a bypass of containment for steam generator k.Lge. Upon implementation of the voltage-based repair criteria, it must be verified that the expected distributions of cracking indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary dose within the current licensing basis.
Data indicate that a threshold voltage of 2.8 volts could result in through-wall cracks long enough to leak at steam line break conditions. Application of the proposed repair criteria requires that the i
current distribution of a number ofindications versus voltage be obtained during the refueling outages. He current voltage is then combined with the rate of change in voltage measurement and i
a voltage measurement uncertainty to establish an end of cycle voltage distribution and, thus, leak i
rate during steam line break pressure differential. He leak rate during a steam line break is further increased by a factor related to the probability of detection of the flaws. Ifit is found that the s
- Significant Hazards Evalettion Page 4 Voltage-Based Repair Cnteria i
potential steam line break leakage for degraded intersections planned to be left in service coupled l
with the reduced allowable specific activity levels result in radiological consequences outside the current licensing basis, then additional tubes will be plugged or repaired to reduce steam line break leakage potential to within the acceptance limit. Thus, the consequences of the most limiting -
design basis accident are constrained to present licensing basis limits, and therefore there is no l
l change to the probability or consequences of an accident previously evaluated.
l 2)
The proposed license amaadmant does not create the possibility of a new or different kind of i
accident from any accident previously evaluated.
i Implementation of the proposed voltage-based tube repair criteria does not introduce any l
l significant changes to the plant design basis. Use of the criteria does not provide a mechanism that l
could result in an accident outside of the region of the tube suppon plate elevations. Neither a l
l single or multiple tube rupture event would be expected in a steam generator in which the repair l
criteria have been applied during all plant conditions. De bobbin probe signal amplitude repair i
criteria are established such that operational leakage or excessive leakage during a postulated l
steam line break condition is not anticipated. Southern Nuclear has previously implemented a
(
l maxunum leakage limit of 140 gpd per steam generator. De R.G.1.121 criterion for establishing j
operational leakage limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. De 140 gpd limit
(
p provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected l
single crack resulting in leakage that is associated with the longest permissible crack length. R.G.
1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest l
permissible crack is exceeded The longest permissible crack is the length that provides a factor of safety of 1.43 against bursting at steam line break pressure differential. A voltage amplitude of approximately 9 volts for typical outside diameter stress corrosion cracking corresponds to meeting i
this 'ube burst requirement at the 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, a typical burst pressure versus through-wall crack length correlation is used below to define the " longest permissible crack" for evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst at 1.43 times steam line break i
pressure differential and steam line break conditions are about 0.54 inch and 0.84 inch, respectively. Normal leakage for these crack lengths would range from about 0.4 gallons per minute to 4.5 gallons per minute, respectively, while lower 95% confidence level leak rates would i
range from about 0.06 gallons per minute to 0.6 gallons per minute, respectively.
j l
An operating leak rate of 140 gpd per steam generator has been implemented. This leakage limit provides for detection of 0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks at the i
lower 95% confidence level leak rates. Thus, the 140 gpd limit provides for plant shutdown prior l
to reaching critical crack iengths for steam line break conditions at leak rates less than a lower 95% confidence level and for three times normal operating pressure differential at less than nominal leak rates.
l Considering the above, the implementation of voltage-based repair criteria will not create the possibility of a new or different kind of accident from any previously evaluated.
3 I
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- Significant Hazards Evaluation Page 5 Voltage-Based Repair Criteria 3)
De proposed license amendment does not involve a signi& ant reduction in margin of safety.
l The use of the voltage-based repair criteria is demonstrated to maintain steam generator tube
_ integrity commensurate with the requirements ofGeneric Letter 95-05 and R.G.1.121. R.G.1.121 describes a method acceptable to the NRC staff for meeting GDC 2,14,15,31, and 32 by reducing the probability of the consequences of steam generator tube rupture. This is accomplished by determuung the limiting conditions of degradation of steam generator tubing, as j
established by inservice inspection, for which tubes with unacceptable crackmg should be removed from service. Upon implementation of the criteria, even under the worst case conditions, the' occurrence of outside diameter stress corrosion crackmg at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions.
i
- The most limiting effect would be a possible increase in leakage during a steam i Neak event.
Excessive leakage during a steam line break event, however, is precluded by verifying that, once i
the criteria are applied, the expected end of cycle distribution of crack indications at the tube l
support plate elevations would result in nurumal, and acceptable primary to e-d=y leakage during the event and, hence, help to demonstrate radiological conditions are 3ess than an appropriate fraction of the 10 CFR 100 guideline.
j i
The margin to burst for the tubes using the voltage-based repair criteria is comparable to that currently provided by existing technical specifications.
i in addressing the combined effects of LOCA + SSE on the steam generato component (as required by GDC 2), it has been determmed that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result oflateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.
l There are two issues associated with steam generator tube collapse. First, the collapse of steam i
generator tubing reduces the RCS flow area through the tubes. The reduction in flow area j
increases the resistance to flow of steam from the core during a LOCA which, in turn, may i
potentially increase Peak Clad Temperature (PCT). Second, there is a potential the partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse or that short through-wall indications would leak at significantly higher leak rates than included in the leak rate assessments.
Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC 4) is applicable to the Farley reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently.
Iow that they need not be considered in the structural design basis of the plant. Excluding breaks
- in the RCS primary loops, the LOCA loads from the large branch line breaks were analyzed at Farley and were found to be ofinsufficient magnitude to result in steam generator tube collapse or significant deformation.
Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley, any flow area reduction is expected to be minimal (much less than 1%) and PCT margin is available to account for this potential effect. Based on analyses' results, no tubes near wedge locations are
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- Significant Hazards Evaluttion Page 6 Voltage-Based Repair Criteria l
t expected to collapse or deform to the degree that secondary to pnmary in-leakage would be l
increased over current expected levels. For all other steam generator tubes, the pos;ibility of
{
maad ey-to-pnmary leakage in the event of a LOCA + SSE event is not significant. In actuality, j
the amount of waad y-to-primary leakage in the event of a LOCA + SSE is expected to be less than that originally allowed, i.e.,500 gpd per steam generator. Furthermore, secondary-to-primary l
in-leakage would be less than primany-to--nadary leakage for the same pressure differential since i
the cracks would tend to tighten under a wnadary-to-primary pressure differential. Also, the presence of the tube support plate is expected to reduce the amount ofin-leakage.
I Addressing the R.G.1.83 considerations, implementation of the tube repair criteria is l
supplemented by 100% inspection requirements at the tube support plate elevations having outside l
diameter stress corrosion cracking indicatioru,, reduced operating leakage limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating probe inspection requirements for the larger indications left in service to characterize the principle degradation
{
mechanism as outside diameter stress corrosion crackmg i
As noted previously, i:nplementation of the voltage-based repair criteria will decrease the number l
of tubes that must be taken out of senice with tube plugs or repaired. The installation of steam i
generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the voltage-based repair criteria will maintain the margin of flow that would otherwise be reduced through increased tube plugging or sleeving.
Considering the above, it is concluded that the propased change does not result in a significant
'l reduction in margin with respect to plant safety as defined in the Final Safety Analysis Report or any bases of the plant Technical Specifications.
CONCl.USION Based on the preceding analysis, it is concluded that using the voltage-based repair criteria in accordance i
with Generic Ixtter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes l
Affected by Outside Diameter Stress Corrosion Cracking," for removing tubes from senice or repairing tubes at Farley is acceptable and the proposed license amendment does not involve a Significant Hazards Consideration as defined in 10 CFR 50.92.
Justification for Changes in Analysis 1
i 4
4
Justification for Changes in Analysis By Southern Nuclear letter dated December 9,1993, a 2 volt interim steam generator tube repair criteria was requested for Farley Unit 1. Subsequent to that letter, a 2 volt repair criteria was approved and has been in are since that time. Based on these submittals, the current accident induced leakage limit is 11.2 gp'n for Farley Nuclear Plant. On August 3,1995, the NRC issued Generic Letter 95-05, Voltage Based Repair Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking. While allowing a permanent alternate repair criteria for outside diameter stress corrosion cracking at tube support plates, the Generic Letter set specific requirements for probability of detection and the inclusion of RPC NDDs in the condition monitoring and operational assessments.
For the alternate repair criteria, the Generic Letter requires that:
a.
the conditional probability of burst in the event of a steam line break should not exceed I x 102,and b.
that dose rates in the event of a steam line break should not exceed the allowable limit determined by the licensing basis of the plant.
Nothing in this submittal is intended to change these requirements associated with probability of burst or
)
dose rates. These requirements provide adequate safety margin for operation of the steam generators.
i 1.0 Modifications To Analysis Methodology in conjunction with the technical specification amendment, several additional modifications are requested for the analysis methodology in support of this revised voltage-based alternate repair criteria (ARC). Each of these changes is referred to in Generic Letter 95-05. These modifications include:
l a.
revised steam line break leakage limit - Generic Letter 95-05, Attachment 1, Section 2.b.4 allows the use of reduced iodine activities for dose rate calculations. The calculation determined that with a primary-to-secondary steam generator tube leakage rate in a steam generator with a steam line break equal to 20 gpm, the dose limits of NUREG-0800 and GDC 19 continue to be met.
b.
application of voltage dependent probability of detection - Generic Letter 95-05,, Section 2.b.1, allows use of an alternative probability of detection function if approved by the NRC.
1 c.
exclusion of a fraction of RPC NDD indications in steam line break leak rate and burst l
probability analyses - Generic Letter 95-05, Attachment 1, Section 2.b.1 allows this exclusion subject to NRC approval.
i
' 2.0 Justifications for Modif ons to the Analysis
' ' Justification for Changes in Analysis Page 2 Voltage-Based Repair Criteria 2.1 Revised Steam Line Break Leakage Limit he calculation to determine the maximum permissible primary-to-secondary leak rate during a steam line break for Farley Units 1 and 2 has been revised. Assumptions ned in the evaluation are provided in Table 2-1. Pnmary changes to the calculation from that submitted by ENC letter dated June 4,1992, are:
a.
The calculation was based on primary coolant iodue 9ctivity for an accident initiated iodine spike of 500 times the equilibrium appearance rate starting from 0.5 pCi/gm and for a pre-accident iodine spike of 30 pCi/gm of dose equivalent I-131.
I
- b. He thyroid dose conversion factors ofICRP-30 were used instead of Regulatory Guide 1.109.
The calculation determmed that with a pnmary-to-secondary steam generator tube leakage rate of 20 gpm in a steam generator with a steam line break, the following acceptance criteria continue to be met:
a.
For a steam line break with an assumed pre-accident iodine spike, the calculated doses will not exceed the guideline values of 10 CFR 100 (25 REM whole-body or 300 REM th>Toid).
- b. For a steam line break with the equilibrium iodine concentration in combination with an assumed accident initiated iodine spike, the calculated doses will not exceed a small fraction of the 10 CFR 100 guideline values (10% of the guidelines, i.e.,2.5 REM for whole body and 30 l
REM for thyroid doses).
c.
The control room operator dosage will be limited to 5 REM whole body or its equivalent to any part of the body (30 REM th>Toid).
1 2.2 Application of Voltage Dependent Probability of Detection (POD)
A voltage dependent POD has been developed based on data from fifteen inspections at eight plants and it is termed probability of prior cycle detection (POPCD). POPCD is obtained by evaluation of all indications found in a given inspection relative to whether they are new indications or indications detected j
and reported at the prior inspection, and it defmes a realistic POD for the prior cycle. Farley-1 EOC-11 and EOC-12 inspections, and Farley-2 EOC-8 and EOC-9 inspections are among the 15 inspections for
]
which POPCD has been established.
With a constant POD, as recommended in GL 95-05, an additional 0.67 of an indication is included in the BOC distribution for leak and burst analyses for each detected indication. POD-driven indications are included in the BOC distribution even if repaired and these indications contribute significantly to the projected leak rates and tube burst probabilities. Since the ARC experience data indicate that POD approaches unity above about 2.5 to 3.5 volts, the larger voltage POD-driven indications are " fictitious" indications in the analyses above a few volts. It is, therefore, appropriate to incorporate a voltage dependent POD in the ARC analyses.
For ARC applications, the important indications are those that could significantly contribute to EOC leakage or burst probability. Rese significant indications can be expected to be detected by bobbin and confinned by RPC inspection. Rus, the population ofinterest for ARC POD assessments is the EOC, RPC confirmed indications that were detected or not detected at the EOCu inspection. EOC, is the end of thejust completed operating cycle and EOCu is the end of the previous operating cycle. POPCD can then be defmed as:
)
1
- Justification for Changes in Analysis Page 3 Vcitage-Based Rcpair Criteris l
EOC, RPC Confirmed and + E O C,i RPC confirmed and Detected at EOC,i Plugged at EOC :
POPCD(EOC,i) =
{Namerator)
+
New EOC, RPC Confirmed l
Indications (i.e., not detected l
at E O C i)
POPCD is evaluated at the EOC i voltage values (preferably from EOC, reevaluation for growth rate or
{-
prior inspection results ifnot reevaluated) since it is an EOC i POPCD assessment. He indications at EOC,i that were RPC confirmed and plugged are included as it can be expected that these indications l
would also have been +*~ *~i and confirmed at EOC.. It is also appropriate to include the plugged tubes for ARC applications since POD adjustments to define the BOC distribution are applied prior to reduction i
of the EOC indication distribution for plugged tubes.
l It should be noted that the above POPCD definition includes all new EOC, indications not reported in the EOC inspection. Ec, new indications include EOC i ndications present at detectable levels but not j
i reported, indications present at EOC,i below detectable levels, and indications that initiated during Cycle
- n. Hus, this definition, by including newly initiated indications, differs from the traditional POD dermition.
Since the newly initiated indications are appropriate for ARC applications, POPCD is an acceptable definition and eliminates the need to adjust the traditional POD for new indications.
Ec above defmition for POPCD would be entirely appropriate if all EOC, indications were RPC inspected. Since only a small fraction of bobbin indications (e.g., < 2.0 volts) are RPC inspected, POPCD could be distorted by using only a few indications in this voltage range. In this case, a more appropriate l
POPCD estimate can be made by assuming that all bobbin indications not RPC inspected would have been RPC confirmed. His definition is applied only for the EOC, indications not RPC inspected since inclusion l
for the EOC i inspection could increase POPCD by including indications on a tube plugged for non-l ODSCC causes. His POPCD can be obtained by replacing the EOC, RPC confirmed by RPC confirmed plus not RPC inspected in the above definition of POPCD.
he POPCD evaluations for inspections performed since 1992 show significant improvement over the earlier assessments which represent the first interim plugging criteria (IPC) inspections. Bobbin data analysis guidelines (Appendix A guidelines) have been revised since the first inspections to reflect the initial IPC experience. hus, it is appropriate to assess POPCD for inspections performed since 1992. Eleven of the fiftecn_ inspections for which POPCD has been evaluated were performed since 1992 including EOC-12 inspection for Farley-1 and EOC-9 inspection for Farley-2.
I Figure 2-1 shows the POPCD data based on over 30,000 indications from 11 ARC inspections performed after 1992. He data shown are based on RPC confirmed plus not RPC inspected indications, and a more detailed assessment of the data is given in Reference 4.1. He nommal and lower 95% confidence limit on the POPCD data is shown in the figure. Also shown is an EPRI POD developed from testing field analysts against a data set ofindications with the definition of"true" indications developed by a group of very experienced NDE analysts. It is seen that the POPCD and EPRI PODS are in very good agreement which supports the adequacy of the voltage dependent POD. Figure 2-2 shows the POPCD distributwn recommended for ARC applicatioc. It was obtained by drawing a continuous curve through the lower 95%
r
' ' Justification for Changes in Analysis Page 4 Voltage-Based Repair Criteria confidence limit at the mid-point of each voltage bin used for the assessment. Tabular data for the recommended POPCD distribution are shown Table 2-2.
It is concluded that a large (over 30,000 indications from 11 inspections) ARC field experience database is available and has been applied to develop a voltage dependent POD. The recommended POD, based on the i
lower 95% confidence limit of the data, is lower than the NRC recommended POD value of 0.6 below about 0.5 volts, increases to 0.9 at 1.2 volts, and approaches unity at 3.5 volts. Flaw population growth increases the need to apply a voltage dependent POD to reduce the number of high voltage " fictitious" indications included in the leak and burst analysis as a result of applying arbitrarily constant and low PODS at higher voltages.
2.3 Fraction of RPC NDD Indications in SLB Leak Rate and Tube Burst Probability Analyses GL 95-05 specifies that all RPC NDD indications are to be included in leak and burst analyses for both the actual indications found in an inspection and the projected EOC indications. The Generic Letter does provide a provision for using a fraction of the NDD indications in the analyses upon further NRC approval.
The inclusion of an RPC NDD indications in the ARC supporting analyses increases the consenatism of the leak and burst results.
Significant field experience has been obtained in tracking ARC inspections relative to the number of RPC NDD indications left in senice that become confirmed _ indications at the next inspection. A summary of these data from ten ARC inspections, including two Farley SG inspections, is given in Table 2-3. It is seen that only 44% of the RPC NDD indications left in senice at Farley-1 became confirmed indications at the subsequent inspection and none of a small population (10 indications) became confirmed at Farley-2.
Industry wide, the subsequent cycle confirmation rate of NDD indications left in senice has ranged from 7% to 60% Sensitivity analyses for Farley-1 indic-that including the operating experience basis for about 50% of the RPC NDD indications in the propted EOC distributions would result in decreases in the SLB leak rate by about 8% and the burst probability by about 10%
l Inclusion of all RPC NDD indications in the leak and burst analyses represents a consenatism in the analysis process that is independent of an increase in the voltage repair limit. It is recommended that a fraction of 50% of the RPC NDD indications left in senice be included in the projected ECC acalyses.
3.0 Conclusion i
l In conjunction with the technical specification amendment, several additional modifications have been requested for the analysis methodo', y. Each of these changes is allowed by Generic Letter 95-05. These modifications include:
I a.
revised steam line break leakage limit; b.
application of voltage dependent probability of detection; and c.
exclusion of a fraction of RPC NDD indications in steam line break leak rate and burst probability analyses.
The analyses contained in Section 2 provide the bases for makmg these changes. Nothing in these requested changes or in this submittal is intended to change these requirements associated with probability of burst or dose rates. These requirements provide adequate safety margin for operation of the steam generators. Consequently, implementation of these analysis changes is justified and allowed by Generic Letter 95-05.
.... ~ -. _ _.. _ _ _ _. _. _
~ '
Justification for Changes in Analysis Page5 Voltage-Based Repair Criteria i
4.0 References 4.1 NRC Generic Letter 95-05, " Voltage Based Repair Criteria for the Repair of Westinghouse Steam i
Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," USNRC Office of Nuclear Reactor Regulation, August 3,1995.
4.2 Draft of Addendum-1 to EPRI Report NP-7480-L, " Steam Generator Outside Diameter Stress Corrosion Cracking at Tube Support Plates - Database for Alternate Repair Criteria," August 1996.
f i
I 4
a h
i 4
1 i
1 l
r 4
0 4
Justification for Changes in Analysis Page 6 Voltage-Based Repair Criteria Table 2-1 2
Major Assumptions Used in Offsite and Control Room Dose Analyses for SG Alternate Plugging Criteria RCS Activity Pre-accident 12 Spike (del n pCi / gm) 30 i
lin 23.3 Ii32 8.38 I
37.2 i33 Iiu 5.59 4
I 20.5 i35
?.
Initial 1 Concentration (DElin Ci/ gm) 0.5 2
for Accidentinitiated1 Spike 2
Accident Initiated Appearance Rate (Ci / sec)
(500 x 0.5 Ci/gm RCS Equilibrium Rate) t lin 0.67 In2 1.265 Im 1.525 liu 1.95 Ii33 1.435 Accident Initiated Fuel Failure (%)
0 Initial Secondary Coolant Activity (DElin pCi / gm) 0.1 Offsite Power, Turbine Condenser Not Available RCS Volume (including CVCS) (ft')
1.071 x 10' Steam Releases (ib. - not including RCS leakage) 5 Intact SG 4.79 x 10 Ruptured SG 9.62 x 10' RCS leak to intact SGs (gpd each) 140 lodine Partition Factor Faulted SG 1
RCS Leakage 1
Intact SG 0.1 Offsite Atmospheric Dispersion (X Q)
FSAR Table 2.3-12
/
Jose Conversion Factors ICRP-30
' ' Justification for Changes in Analysis Page 7 Voltage-Based Repiir Criteris Table 2-1 continued Control Room Model Volume (ft')
1.14 x 10' 5
Normal HVAC Intake Rate (cfm) 1.35 x 10 Duration (min.)
1 Pressurization Mode HVAC 2
Filtered Intake (cfm) 4.5 x 10 Intake Filter Efficiency (%)
99 Unfiltered Intake (cfm) 10 Recirculation Flow Rate (cfm) 2.7 x 10' Recirculation Filter Efficiency (%)
95 Atmospheric Dispersion Factor (y, / Q - sec/m')
3.28 x 10
l I
J
Io 1
Justification for Changes in Analysis Voltage-Based Repair Criteria Page 8 1
i Table 2-2 Comparison of Recommended P00 with EPRI POO Voltage EPRI Recommended Bin POO P00 0.1 0.30 024 0.2 0.34 0.34 0.3 0.49 0.44 y
O.4 0.6/
0.63 0.5 0.62 0.62 0.6 0.66 0.67 O.7 0.71 0.73 0.8 0.76 0.77 0.9 0.80 0.81 1
0.83 0.43 1.2 0.90 0 86 1.4 0.93 0.91 1.6 0.96 0.92 1.8 0.96 0.93 2
0.964 0.94 3
1.00 0.98 3.5 1.00 1.00
\\
4 Justification for Changes in Analysis Page 9 Vo'tage-Based Repair Criteria Table 2-3 RPC Conllrmation Rates for RPC NDD ladicaelons in the Last inspect 6on Compoolte Data for All Steam Gonoratofs in Plant RPC NDO in First snepecton - Second Inspecmon Date First Seoend s 1 volt W in 2"Inapar+wwi
> 1 watt Solein in 2" %
As~ % in2" W l
Plant inspecmon inspecean No. of No of Percent No. of No. of Percent No. of Noet Percent Inecemone tw.nuwie gnecedone Inscamens im tr*=awww Cyde1 year CycesIyear RPC RPC RPC RPC RPC RPC l
e Cenernwa ineeemos Canamnes e
Conannes Farter-1 O
O 84 37 44 0 %
84 37 44 0 %
O O
M 0
m W
0 N
j Fartsy-2 1 3 GS i
Plant D-1 0
0 2
1 50 0 %
2 1
50 0 %
'3 r
OC 10 E 11 10 6
30 0%
10 6
60 0 %
Plant P-1 0
0 1995 1998
.t Plant P-1 21 16 85.7 %
80 38 42.7 %
110 56 50 9 %
i Peent F 0
0 0
0 0
0 EOC 19 EOC30 1994 1988 Plant R 08 0
0%
46 8
17.8%
113 8
71%
3 1
Plant AA-1 21 10 47 0%
177 71 40 1%
198 81 40 9%
0C4 EOCSA
,p ggg EOC6 EOC7A Pient As-i 11 3
27.3%
3 10 625%
27 13 de1%
p Plant As-1 30 9
30 0 %
3 0
00%
33 9
27 3%
I
^
4 Justification for Changes in Analysis Page 10 Voltage-Based Repair Criteria Figure 2-1 C.embleed POPCD Evelemtion for 11 Post '92 leapectices Force sesed.e arc c.enremed Flus Not Inspected Indecet6ees r
__,._m.-
--- ---__=._.
- 7 1.0 1
L. -. -. -..g--
e '- - - - d..
0.9 n--s s'
4 >-~&i 0.8 1r a--yu 0.7
/
f
)
s.
0.6
--"~'-"
)
- 0.5 f-Data from 11 Inspechons i i -r- - W i I
j 0.4 ---P j
~~
/
E3' -5=) i
- e-EPRI POO 0.2
-*-- 95% Lower Confklence Limit 0.1 0.0 0
0.5 1
1.5 2
2.5 3
3.5 Bobbie Anspiltede
...-~.. -. _. - - _
~
e L
PageIi Justification for Changes in Analysis s
Voltage-Based Repair Criteria i
Figure 2-2 CoeWeed FOPCD Evalenties for II Pose-12 * :, cir POPCD h om RFC Coenrased Plas Not Isopected Ind6 cat 6ees i
l 1.0
- ,x-e#
y-s 0.9 0.8
/
0.7 s
8 30.6 i
?
95% Lower Con 6dence Limit for e
o q_$
D
/
Data from 11 Inspechons 5
/
/
j 0.4 e
/
s.
p Recommended POD O.3 -i 4
0.2
--x-EPRI POD 0.1 i
0.0 ::
t 0
0.5 1
1.5 2
2.5 3
3.5 i
BobMe Amplitude
.