ML20114B409
| ML20114B409 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/24/1992 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20114B408 | List: |
| References | |
| NUDOCS 9208280055 | |
| Download: ML20114B409 (20) | |
Text
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l ATTACHAENT 2 Marked-up Technical Specifications 9208280055 920824 DR ADOCK 05000413 PDR
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
2)
Tubes in those areas where experience has indicated potential problems, and 3)
A tube inspection (pursuant to Specification 4.4.5.4a.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 be selected and subjected to a tube inspection.
c.
For Unit 1, in addition to the 3% sample, all tubes for which the alternate plugging caiteria has been previously applied shall be inspected in the tubesheet region.
d.
The tubes selected as the second and third samples (if required oy 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 j
imperfections were previously found, and l~
2)
The inspections include those portions of the tubes where dI imperfections were previously found.
p
, e.
Q The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between I
5% and 10% of the total tubes inspected are i
degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
CATAWBA - UNITS 1 & 2 3/4 4-13 Amendment No. 47 (Unit 1)
Amendment No. 40 ' Unit 2) l
4 CJG 18 ' 92 16 16-FROM TECH $PECS TO 88338313191-PAGE,ccareia 2
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G ad i INSERT A
- e. # haplementation of the interim steam generator tube /n2bc support plare elevation plugging limit i
requires a 100% bobbin probe iWan for all hot leg tube support plate intersections aM all cold leg intersections down to the lowest cold leg tube support plate with outer diamerer stress corrosion cracking (OD SCC) indications. An inspection using the rotating pancake
/
- coil (RPC) probe is required in order to show operability of tubes with flaw like bobbin coil
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signal amplitudes greater than 1.0 volt but!ess than@ volts. For tubes'that will be adminieratively plugged no RPC inspection is requir The RPC results are to be evaluated to establish that the p ' pal indications can be char ized as OD SCC.-
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- AUG 18 '92 16:16 FRCri TECH SPECS TO 88038313191 PGE. 005<014 REACTOR C00 TANT SYSTEN SURVElt. LANCE REQUIREME,NTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the folicwing frequenciez:
The first inservice inspection shall be perfonned after 6 Effective a.
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, not including the preservice inspection, result in all inspection results falling into the C-1 catigory 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 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' 4.4.5.3a.; the interval may then be extended to a maximum"of once per 40 months; and Additional, unscheduled inservice inspections shall be perfonned 'on c.
each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subseqwnt to any of the following conditions:
1)
Reactor-to-secondary tubes leak (not including leaks originating from tubt-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A loss of-coolant accident requiring actuation of the Engineered Safety Features, or 4)
A main steam line or feedwater line break.
x Fbrum+1
.t PC.
4 Tubes in which the tube support plate elevation)1ugging limh have been applied d.
j shall be inspected during all future refueling outages.
l CATAVBA - UNITS 1 & 2 3/4 4-14
AUG 19 '92 16:17
- ROM TECH SPECS TO 89038313191 PfG.006/014 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 or sleeve from that required by fabrication l
drawings or specifications.
Eddy current testing indications below 20% of the ^ominal tube or sleeve 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 or outside of a tube or sleeve; I
3)
Degraded Tube means a tube or sleeve containing imperfections greater than or equal to 20% of the nominal tube or sleeve wall
}
thickness caused by d4 gradation; I
4)
% Degradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation; i
5)
Defect means an imperfection of such severity that it exceeds 3
the repair limit.
A tube or sleeve containing a defect is defective; j
6)
Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving.
It also means the imperfection depth at or beyond which a sleeved tube shall be plugged.
to 40% of the nominal tube or sleeve wall thickness.The repair limit is equal For Unit 1, this definition does ne*, apply to the region of the tube subject to the alternate tube plugging criteria.
If a tube is sleeved due to degradation in the F* distance, then e
any defects found in the tube below the sleeve will not necessi-tate plugging.
(\\
b The Babcock & Wilcox process described in Topical Report BAW-
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'45-2045(P)-A will bs used for sleeving.
7)
Unserviceable describes the condition of a tube if it leaks or contains a defect 1arge 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 4.4.5.3c., above; 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the INSERT __L-u-bend to he top support or the cold les;-
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C.
E RE.
ATAWBA - UNITS 1 & 2 3/4 4-15 Amendment No. 84 (Unit-1) wn%,,, % 7o ns.u e.
AUG 10 '92 16327
'FROM TECH SPECS TO 98038313191 PAGE.007 014 R_MQORCOOLANTSYSTEM
_SURVEIUANCE REQUIREMENTS (Continued) 9)
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 prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10)
Tube Roll Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice ex'ists between the outside diameter of the tube and the tubesheet.
11)
F" Distance is the minimum length of the roll ex;:anded portion of the tube which cannot contain any defects in order to ensure the tube does not pull out of the tubesheet.
The F* distance is 1.60 inches and is measured from the bottom of the roll expansion transition or the top of the tubesheet if the bottom of the roll expansion is above the top of the tubesheet.
Included in this distance is a safety factor of 3 plus a 0.5 inch eddy current vertical measurement uncertainty.
12)
Alternate __ tube plugging criteria does not require the tube to be removed from service or repaired when the tube degradation exceeds the repair limit so long as the degradation is in that portion of the tube from F* to trie bottom of the l
tubesheet.
This definition does not apply to tubes with INSER i.
degradation (i.e., indications of cracking) in the F*
g i
distTnce.
WHE b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the repair limit and all tubes containing through-wall cracks) required l
by Table 4.4-2.
For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria and do not have to be plugged 4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes repaired in each steam i
generator shall be reported to the Comission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted'to the Comission in a Special lleport pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall includa:
1)
Number and extent of tubes inspected, CATAWBA - UNITS 1 & 2 3/4 4-16 Amendment No. 84 (Unit 1)
Amendent Nn 78 to.a e
g.g2 99 N O TECH SPECS TO 88038313191 P N E.00a 014 6 Und 1
^ Also, this definition does not apply for tubes experiencing outer diamets stress corro cracking confirmed by bobbin probe inspection to be within the thickness of the tube supp plates. See 4.4.5.4 a 13 for the plugging linut for use within the thickness of the tube su plate.
INSERT C
% voM L (rec.)
A For a tube in which the tube support plate elevation interim pluggmgiimit has been appli the inspection will include all the hot leg intersections and all cold leg intersections down to and iticluding, at least, the level of the last crack indication foR wd/<// TNE ivreR/m fuso/*G CRITERI A LitutIT IS To Tbl NWED-INSERT (n Q_
en 13.
De Tube Succort Plate Interim Plurein/Ligiis for disposition of a steam generator tube for continued service that is experienc ng outer diameter initia:ed sinss corrosion cracking confined within the thickness of tie tube support plates. For application of the tube support plate intaim plugging limit, the tube's disposition for continued service will be b: sed upon standard bobbin probe signal amplitude of flaw like indications. De plant specific guidelines used for all inspections shall be amende as appropriate to accommodate the widitional information needed to evaluate tube support plate signals with respect to the voltage / depth paramaters as specified in Specific.aticn 4.4.5.2. Pending incorporation of the voltage verification requirement in
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ASME standara verificatio
,1 ASMS stand -
ibrated laboratory standard will be utilized i 4:e-wba Unit Ilh= ;;mer Lep consistent voltage normal' r
1.
A tube can remain in service if the signal amplitude of a crack indication is less than or equti to 1.0 volts, regardless of the depth of tube wall penetration, if, as a result, the projected end of cycle distribution of y'g
[iA/0wDU ofEMhXL indicaticus is verified to result '
rimary to secondary leakage less th. 04-
- A"#.
. The basis for determming e leak rates from the projected C
LEAKAGE) distribution is provided in SECL-92-282.
TOTAL 2.
A tube can remain in scryice with a bobbin coil signal amplitude greater than 1.0 volt but iess than inspection does not d volts provided a rotatmg pancake coil (RPG adation.
3.
Indications of degradation vith a flaw type bobbin coil signal amplitude D"
greater than olts will be plugged or repaired.
2.. b Certain tubes, as i4cDuified in SECL-92-20, will be excluded from application of th Interim PluggingT,imit(Critenh it has been daermmed that these tubes may collapse or deform following a postulated LOCA + SSE Event.
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qf, gg.92 16 19 FROM TECH SPECS TO 90038713191 NE. NW I
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2)
Location and percent of wall-thickness penetration for indication of an imperfection, and each 3)
Identification of tubes repaired.
For Unit 2, results of steam generator tube inspections, which fall c.
into Category C-3 shall be reported in a Special Report to the Coensission pursuan,t to Specification 6.9.2 within 30 days and prior to resumption of plant operatien.
This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to pmvent recurrence d.
For Unit 1, tho results of inspections for all tubes for which the alternute tube plugging criteria has been applied shall be reported to the Nuclear Regulatory Commission in accordance with 10 CFR 50 4 prior to restart of the unit followire the inspection.
shall include:
This report 1)
Identification of applicable tubes, and 2) location and size of the degradation.
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For Unit 1, the results of inspecuons performed undar 4.4.5.2 e..
for all tubes in wbkh the tube support plate elevations ira %
plugging limit has been applied shall be reported to the Commation following the inspection and prior to the resumption of plant I
operanon. Tbc report shall include:
1.
N Listing of applicable tubes.
2.
Locatico (applicable intersections per tube) and extear of degradation (voltage).
b 1
CATAWBA - UNITS 1 & 2 3/4 4-16a Amendment No. 34
(_ Unit 11_
g 1g.92 16:19 FROM TECH SPECS TO 880 E 13191 N #N REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDIT_ ION FOR OPERATION 3
3.4.6.2 Reactor Coolant Systa leakage shall be limited to; No PRESSURE BOUNDARY LEAKAGE, a.
c b.
1 gpm UNIDENTIFIED' LEAKAGE,
@ gp..
and @ gallons per day through any one steam genert. tor,a t c.
fibio -
A V
d.
10 gpa IDENTIFIED LEAKAGE from the Reactor Coolant System, 40 gpa CONTROLLED LCAKAGE at a Reactor Coolant System pressure of e.
2235 1 20 psig, and f.
I gpa leakage at a Reactor Coolant System pressure of 2235 t 20 psig from any Reactor Coolant System Pressure Isolation Valve fpecified in Table 3.4-L APPLICABILO: MODES 1, I., 3, and 4.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANOBY a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any Reactor Coolant Systen leakage greater than any one of the above limits, excluding PRESSURE BOLHDARY LEAKAGE and leakage from Reactor Coolant Systee Pressure Isolation Valvte.
rate to 'vithin licits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANOBYreduce the leakage within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l With any Reactor Coolant System Pressure Isolation Valve leakage c.
greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STAMOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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' AUG 19 ' 92 16:19 FROM TECH SPECS TO 88038313191 PAGE.011/014 REACTOR C00(. ANT _$YSTEM e
BASES STEAM GENERATOR $ (Continued) been isoleted due to excessive seat leakage and except for limit wnere the PORY and/or block valve is c Msed because of tes evolution is coverto by an approved procedure.
challenges to the code safety valves for overpressurization events. 5) ManualT
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control of a block valve to isolate a stuck-open PORV.
Testing of the PORVs includes the emergency N2 apply ros the Cold Leg Accumulators. This test demonstrates that the va m s in the supply line operate satis proper PORY operation.
3/4.4.5 STEAM GENERATORS The Surysillance Requirements for inspection of the stems generator tut 4s System will be maintained. ensure that the structural integrity of this portion of the Reactor Co The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
tain surveillance of the conditions.cf the tubes in the event.t evidence of mechanical dage or progressive degradation due to design, manu-facturing errors, or inserv' ce conditions that lead to' corrosion.
inspection of steam generator tubing also provides a means 6f characterizing the Inservice nature at4 cause of any tube degradatica so that farrwetive measures can be taken.
Topical Report 8AW-2045(P)-A will be used.The.84W process (or metho erator sleeves is also required to ensure RC3 integrity.Insen 44 inspeer M of steam of eddy current testing, therefore, special inspection metho Because the sleeves method is described in Topical Neport BAtF2045(P)-A with supporting validation A
data that demonstrates the inspe:tability cf the sleeve and As mits to validate the adequacy of any systas that is used for periodic inservice Power Company, implement testing methoas as better met validated for commercial use.
coolant will be asintained within those Desistry limits found to negligible corrosion of the steam 3enerator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
tien would be limited by the limitation of steam generator tree leakage betw the Reactqtfoolant System and the Secondery. Coolant System (reactor-to-seconda leakage 4 900 gallons per day per steam generator.
secondary leakage less than this limit during oper)ation will have an adequateC strgin of Jafety te withstand the ' loads imposed during norsel opeestion and by postulated acc mentsv secondary g g j p srating plants have demonstrated that reactor-to-Ms L,
ator can readily be detected of this b.biWrequfreYant3Fut3FMFChrnen.L_
Laakage in excess which the leaking. tubes will be located and repaired.
inspection, during
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CATAMA - UNITS 14 2 8 3/4 4-3 1
Amendment No.95(Unit 1)
Amendment No.89(Unit 2)
RUG is '92 1o 20
%MN REACTORCOGLANTSYST@
t BASES
_ STEAM GENERATORS (Continued) l Wastage-type defects are unlikely with proper chemistry treatment of th secondery coo' ant.
be fou3d during scheculed insarvice steam generator tube examinations e
40% of the tube nominal wall thickness.will be required for all tubes witn im Repair under the alternate tube plugging criteria do not have to be repair 6d.For U tive steam generator tubes can be repaired by the installation of sleeves which Defec-spara the area of degradation,.and serve as a replacement pressure boundary for the degraded portion of the tube, allowing the tee to remain in service generator tebe inspections of operating plants have demonstrated the capability Steam 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 Commit.ston pur-suant to Specification 6.9.2 prior to resuretion of plant operation.
cases will be considered oy the Commission on a case-by-case basis a'nd may Such result in a requirement for analysis, laboratory exasinations ecoy current inspection, and revision of the Technical Specifice kas, if
, te91, additional necessary.
If a tube is sleeved due to degradation in the F* dis unce, then any defects in the tube below the sleeve will remain in service without repair.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAXAGE DETECTION SYSTEMS
! to n.onitor and detect leakage from the reactor coolant p l
Guide 1.45, " Reactor Coolant Pressure Bouneary Laak j
May 197.1.
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Tubes expenencing outer dianmar stress corrosbu cracbag widdn the thickness of the utbe i
support plass are plugged oc repaired by the critenon of 4.4.5.4.a.13.
i CATAWBA - UNITS 1 & 2 B Va 4-3a Amendment No 95 (Unit 1) laendment No.89 (Unit 2)
.nUG 19 '92 16:21 FROM-TECH SPECS TO 88038313191 PAGE.013/014 REACTOR COOLANT SYSTEM BASES __
3/4.a.6.2 OPERATIONAL LEAKAGE be indicative of an impending gross failure of the pressure bo 1
placed in COLD SHUTDOWN.the presence of any PRESSURE B0UNDARY LEAK Therefore, Industry experience has shown that while a limited amount of leakage is expected from the Reactor Coolant Syster>, the unidentified portion of this 1
leakage can be reduced to a threshold value of less than 1 gpe.
hold value is su'fficiently low to ensure early detec This thres-of additicnal leakage.
The total steam generator tub le limit
$9pmforallsteam generators not isolated from the eactor Coolant Sy te# ensure; that the dosage centribution from the tube leakag Part 100 dose guideline values ip M gevent of e,jther a steam generator rupture or steam line break.
Tip gpm limi ML used in the analysis of these acclunts.
nsistent with the' assumptions generator ensures that steam generator tubeTh '$ p leakage limit per steam event of a main steam line rupture or under LOCA conditionstetJtity is maintained in the amount of leakage frcm known sources whose presence wil the detection of UNIDENTIFIED LEAKAGE by the Leakage. Detection Systems supplied to the reactor coolant pump seals exceeds 40 valve in the supply line fully open at a nominal Reactor Coolant System pres-sure of 2235 psig.
safety injection flow will c.ot be less 'than assumed in the safety The 1 gpm leakage f, rom any Reactor Coolant System pressure isolation val is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-for a substantial length of time, verification of valve integri Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be teste:d periodically to ensure low prota of gross failure.
valves provide added assurance of valve integrity thereby n
ability of gross valve failure and consequent intersystem LOCA.
a portion of the allowed limit.the pressure isolation valve is :9ENTIFIED LEAXAGE Leakage from CATAWBA - UNITS 1 & 2 B "/4 4-4 c
Amendment No. 84_ (Unit 1) w w.
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_ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ ~.
fuJG 10 '92 16881 NM a
REACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEMIST.RY.
t The limitations on Reactor Coolant System chemistry, ensure that corrosion Coolant System leaxage or failure due to stress corrosion.of the Maintaining the chemistry within the Stea$-State Limits provides adequate corrosion protection j
to ensura the structural integrity of the Reactor Coolant Systes over the life of the plant.
fluoride limits are time and temperature dependent.The associated effect operation may be continued with contaminant concentration levels in excess the Steady-State Limits, up to the Transient Limits, for the specified Ifaited" i
of the Reactor Coolant-System. time intervals without having a significa The time interval permittirq continued operation wi' thin the restrictions of the Transient Limits provides time for taking correc tive actions to restore the contaminant concentrations to within the Steady-Stat Limits.
j in excess of the limits will be detected in sufficient time action.
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3/4.4.8 SPECIFIC ACTIVITY 04 I
The limitations on the specific activity of the reactor coolant ensure that the resulting 2-t.our doses at the SITE BOUNDARY will not exceed an a priately small fraction of Part 100 generator tube rupture accident in doso guideline values following a steam conjunction _wi 1
prismry-to secondary steam generator leakage rate nymped stea@ state the limits on specific activity represent limits The values for uation by the NRC of typical site locations.
a parametr in that specific site parameters of the Catawba site, such as S tive 4
location and meteorological conditions, were not considered in this evaluation 1
i time periods with the reactor cocitat's specific activity gre 1.0 microcurie /gran DOSE EQUIVALENT I-131, but within the allowable limit j
shown on Figure 3.4-1, accome Aates possible iodine spiking phenomenon which i
may occur following changes in THERMAL POWER.
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CATAWBA - UNITS _1 &_2 B 3/4 4-5 1
Amendment No. 25 - (Unit 1. ).
. %%aw
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O ATTACHMENT 3 i
i No Significant Hazards Consideration I
and i
j Environmental Impact Statement t
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NO SIGNIFICANT llAZARDS 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 a*id found not to: 1) involve a dgnificant increase in the probability or consequences of an acudent previously evaluated; or
- 2) create the possibility of a new or different kind of accident from any acci&nt previously evaluated; or 3) involve a signincant reduction in margin of safety.
Confonnance of the proposed amendment to the standards for a detennination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
1)
O,eration of Catawba Unit 1 in accordance with the proposed license amendment does not ~olve a significant increase in the probability or consequences of an accident previously evaluated.
Testing of model boiler specimens for free span tubing (no tube support plate restraint) at room temperature conditions show burst pressures in excess of 5475 psi for indications of outer diameter stress corrosion cracking with voltage measurements as high as 11 volts (Reference 1).
Burst testing perfonned on pulled tubes from Catawba Unit I with up to a 1.5 volt indications show measured burst pressures in excess of 4800 psi at room temperature. Correcting for the effects of temperature on material propenies and minimum strength levels (as the burst testing was done at room temperature, tube burst capability signiGcantly exceeds the R.G.1.121 criterion requiring the maintenance of a margin of 3 times normal operating pressum differential on tube burst. The 3 times nonnal operating pressure differential for the Catawba Unit I steam generators corresponds to 3750 psi. Based on the existing data base, this criterion is satisfied with 3/4" diameter tubing with bobbin coil indications with signal amplitudes less than 4.1 volts, regardless of the indicated depth measuremem. This stnictural limit is based on a lower 95%
con 0dence level limit of the data. A 1.0 volt plugging cri'.erion compares favorably with the stnictural limit considering the calculated growth rates for ODSCC within the Catawba Unit I steam generators.
Considering a voltage increase of 0.58 volts, and adding 20% NDE uncertainty of 0.2 volts (90% Cumulative Probability) to the interim plugging criterion of 1.0 volts results in an EOC voltage of 1.78 volts. The growth rate used to detennine the projected EOC voltage is based on the review of growth rates for 541 TSP intersections.
These indications were selected by Duke Power Company based on their largest amplitudes from the original analyses. The 541 indications were made up of 90, Il7,197, and 137 from steam generators A, B, C and D, respectively. This end of cycle voltage compares favorably with the Stnictural Limit 4.1 volt. The corresponding safety margin to the tube structural limit at end of cycle 7 upon implementation of the 1.0 volt steam generator tube interim plugging limit is 2.3 volts. The necessary plugging limit to meet tube structural limits is 2.5 volts.
Only three indications of ODSCC have been reported to have operating leakage - all three have been in European plants. No Geld ieakage has been reported at other plants from tubes with 1 of 6 6
l indications with a voltage level of under 6.2 volts (from 3/4" tubing). Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary to secondary leakage and steam release to the environment are: Feedwater System Malfunction, Loss of External Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Uncontrolled Single Rod Withdrawal at Power, Major Secondary System Pipe l
Failure, Steam Generator Tabe Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. In suppon of implementation of the interim plugging i
criterion, it has been determined that the distribution of cracking indications at the tube suppon plate intersections at the end of cycle 7 are projected to be such that primary to secondary leakage would result in site boundary doses within a small fraction of the 10 CFR 100 guidelines.
l Monte Carlo analyses methods are used to calculate the potential SLB leakage at the EOC-7 at Catawba Unit 1. The Monte Carlo analyses methods utilize the distributions for indications left j
inservice, NDE uncenainties, voltage growth and SLB leak rate. The methods account for the tails of the distribution and yield eddy current voltages with an associated probability of i
occurrence and the cumulative probability of EOC voltages. The SLB teak rates applied to the Monte Carlo voltage distribution are 0.0 gpm for volts less than or equal to 1.8 volts,1 liter /hr for 1.8 to 3.5 volts, and 10 liter /hr for greater than 3.5 volts. Applying these leak rates to the projected EOC voltage distribution leads to a projected SLB leak rate of 0.54 gpm for steam generator D, the most limiting steam generator (3492 TSP elevation indications). The 0.54 gpm SLB leak rate compares favorably with the accident analyses assumptions of 1.0 gpm in the j
affected steam generator identified in Table 15.3 of the Catawba Unit i Safety Evaluation l
Repon. The projection indicates a maximum EPC-7 of 3.1 volts (90% cumulative probability).
The analyses yields a negligible likelihood of a tube exceeding the 3.5 volt threshold for a 10 liter /hr SLB leak rate.
i Upon application of the interim plugging criterion, only a negligible increase in leakage above i
normal operating leakage would be expected during plant transients, other than steam line break, which have lower peak differential pressures.
i Therefore, as steam generator tube burst capability and leaktightness during Cycle 7 operation following implementation of the proposed 1.0 volt interim plugging criterion remains consistent with the current licensing basis, the proposed amendment does not result in any increase in the probability or consequences of an accident previously evaluated with the Catawba Unit 1 FSAR.
s 2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the proposed interim tube support plate elevation steam generator tube plugging criterion does not introduce any significant changes to the plant design basis. Use of the criterion does not provide a mechanism which could result in an' accident outside of the region of the tube suppon plate elevations; no ODSCC is occurring outside the thickness of the tube suppon plates. A tube rupture event would not be expected in a steam generator in which 2 of 6
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the plugging criterion has been applied (during all plant conditions).
Upon application of the interim plugging criterion, no primary to secondary leakage during nonnal operating is anticipated during all plant conditions due to degradation at the tube support plate elevations in the Catawba Unit I steam generators. However, additional conservatism is built into the operating leakage limit with regard to protection against the maximum permissible single crack length which may be achieved during Cycle 7 operation due to the potential occurrence of through wall cracks at locations other than the tube support plate intersections.
Specifically, Duke Power Company will implement a maximum leakage rate limit of 150 gpd i
(0.1 gpm) per steam generator to help preclude the potential for excessive leakage during all i
plant conditions. The currently proposed Cycle 7 Reload Technical Specification limits on primary to secondary leakage at operating conditions is a maximum of 0.5 gpm (720 gpd) for all steam genemtors, or, a maximum of 200 gpd for any one steam generator. The R.G.1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture.
The 150 gpd limit should provide for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest pennissible crack length. R.G.1.121 acceptance criteria for establishing operating leakage limits are based on leak-before break considemtions such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of 3 against bursting at normal operating pressure differential. A voltage amplitude of 4.1 volts for typical ODSCC corresponds to meeting this l
tube burst requirement at a lower 95 % uncertainty limit on the burst correlation, Alternate crack morphologies can correspond to 4.1 volts so that a unique crack length is not delined by the burst pressure versus voltage correlation. Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the " longest pennissible crack" for evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst at 3 times normal operating pressure differential and SLB conditions are 0.48 inch and 0.76 inch, respectively. Nominal leakage for these crack lengths would range from about 0.10 gpm to 3 gpm, respectively, while lower 95% confidence level leak rates would range from -about 0.015 gpm to 0.4 gpm, respectively. A leak rate of 150 gpd will provide for detection of 0.40 inch long cracks at nominal leak rates and 0.60 inch long cracks at the lower 95% confidence level leak mtes.
Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for SLB 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.
Application of the 1.0 volt interim steam generator tube plugging criterion at Catawba Unit I is not expected to result in tube burst during all plant conditions during Cycle 7 operation. Tube burst margins are expected to meet R.G.1.121 acceptance criteria. The limiting consequence of the application of the interim plugging criterior. is a potential for primary to secondary 3 of 6
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leakage of approximatelv 0.54 gpm. This amount of leakage does not result in unacceptable radiological conseq e No unacceptable leakage is anticipated at nonnal operating or RCP locked rotor conditions. Therefore, as the existing tube integrity criteria and accident analyses assumptions and results continue to be met, the proposed license amendment does not create the l
possibility of a new or different k.ind of accident from any previously evaluated.
l 3)
The proposed license amendment does not involve a significant reduction in margin of safety.
i Based on the analysis which shows the new leakage values proposed and the lealage j
characteristics expected during accidents creating high differential pressures across the steam i
generator tubes (main steam line break) new dose analyses were nm to detemiine offsite dose 1
consequencec. A new analysis of the Main Steam Line Break accident using pre-existing leakage's of 0.1 spm per steam generator and leakage growth of 1.1 gpm in the faulted l
generator detennined that the EAB and Low Population Zone doses remain well within 10% of the allowed 10 CFR100 values of 25 Rem whole body and 300 Rem thyroid. The most restrictive dose analysis is the Reactor Coolant Pump Locked Rotor accident which requires that l
total steam generator leakage remains less than 0.7 gpm. This is a new analysis which has been submitted.o suppon Unit 1 Cycle 7.
This accident does not create excessive differential j
pressure conditions across the steam generator tubes and by limiting the initial allowed primary to secondary leakage to 0.4 gpm total,10% of 10 CFR100 dose limits are again not exceeded.
Reruns of the above accident dose analyses show that there is no significant ir. crease in dose consequences.
4 The use of the voltage based bobbin probe interim tube support plate elevation plugging criterion at Catawba Unit 1 is demonstrated to maintain steam generator tube integrity commensurate with the criteria of Regulatory Guide 1.121. R.G.1.21 describes a method acceptable to the NRC staff for meeting GDCs 14,15, 31, and 32 by reducing the probability or the consequences of steam generator tube ruptum. This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by inservice inspection, for which tubes with imacceptable cracking should be removed : vm service. Upon implementation of the criterion, even under the worst case conditions, the occurrence of ODSCC at the tube support plate elevations is not expected to lead to a steam generator tube mpture event during normal or faulted plant conditions. The ene of cycle distribution of crack indications at the tube suppcrt plate elevations is calculated to result in minimal primary to secondary leakage during all plant conditions and radiological consequences are not adversely impacted.
In address the combed effects of LOCA + SSE on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become defonned as a result of lateral loads at the wedge supports at the periphery of the plate due to the combined effects of the LOCA rarefaction wave and SSE loadings. Then, the resulting pressure differential on the defonned tubes may causa some of the tubes to collapse.
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There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS flow-area through the tubes. The reduction in How area increases the resistance to ficw of steam from the core during a LOCA which, in turn, may-potentially increase Peak Clad Temperature (PCT). Second, there is a potential that panial through-wall cracks in tubes could progress to through-wall cracks during tube defonnation or collapse.
Analyses results show that for the Catawba Unit I steam genemtors several tubes near wedge locations may signincantly defonn or collapse and secondary to primary inleakage may result.
These tubes have been precluded from application of interim plugging criterion (Reference 3).
For all other steam generator tubes, the possibility of secondary to primary leakage in the event of a LOCA + SSE event is not significant. In actuality, the amount of secondary to primary leakage in the event of a LOCA + SSE is expected to be less than that associated with the application of this criterion, i.e.,150 gpd per steam generator. Secondary to primary inleakage would be less than primary to secondary leakage for the same pressure differential since the cracks would tend to close under a secondary to primary pressure differential. Additionally, the presence of the tube support plate is expected to reduce the amount of in-leakage.
Addressing R.G.1.83 considerations, implementation of the bobbin probe voltage based interim tube plugging criterion of 1.0 volt is supplemented by: enhanced eddy current inspection guidelines to provide consistency in voltage nonnalization, a 100% eddy current inspection sample size at the tube support plate elevations, and rotating pancake coil inspection requirements for the larger indications left inservice to characterize the principal degmdation as ODSCC.
As noted previously, implementatiori of the tube support plate elevation plugging criterion will decrease the number of tubes which must be repaired or taken out of service by plugging. Tne 1
installation of steam generator tube plugs or sleeves reduces the RCS flow margin. Thus, implementation of the alternate plugging criterion will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.
Based on the above, it is concluded that the proposed license amendment request does not result in a significant reduction in margin with respect to plant safety as defined in the Final Safety Analysis Repon or any BASES of the plant Technical Specincations.
CONCLUSION Based on the preceding analysis, it is concluded that using the TSP elevation bobbin coil probe voltage-based interim steam generator tube plugging criterion for removing tubes from service at Catawba Unit 1 is acceptable and the proposed license amendment does not involve a Significant Hazards Consideration Finding as defined in 10 CFR 50.92.
The proposed Technical Specification change has been reviewed against the criteria of 10 CFR51.22 for environmental considerations. As shown above, the proposed change does not 5 of 6
4 involve any significant hazards consideration,-nor increase the types and amounts of efnuents that may be released offsite, nor increase the individual or cumulative occupational radiation exposures, Based on this, the proposed Technisal Specification change meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement, s
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