ML20093M531
| ML20093M531 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/24/1995 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20093M513 | List: |
| References | |
| NUDOCS 9510260307 | |
| Download: ML20093M531 (11) | |
Text
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d Docket No. 50-336 B15399 Attachment 3 ,
Millstone Nuclear Power Station, Unit No. 2 +
Proposed Revision to Technical Specifications Steam Generator Surveillance Requirement Extension Marked-up Pages e
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e i
October 1995 9510260307 951024
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( SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection sha11'be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspect s shall be performed at intervals of not less than 12 than a 24 calendar months after the previous inspect . two l consecutive inspections following service und nditions, not including the preservice inspection, result all inspec-tion results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the, inspection interval may be extended to a maximum of once per 40 sonths.
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-6 at 40 month intervals fall into 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 inspec-
).
tions satisfy the criteria of Specification 4.4.5.1.3.a; the 1 interval may then be extended to a maximum of once per 40 V months.
i
- c. Additional, unscheduled inservice inspections shall be performed i on each steam generator in accordance with the first sample
! inspection specified in Table 4.4-6 during the shutdown subsequent i to any of the following conditions:
- 1. Pri$y-to-secondarytub s(notincludingleaks
- originatingfromtube-to-tubesheetwelds)inexcess i
of the limits of Specification 3.4.6.2. I
! 2. A seismic occurrence greater than the Operating Basis
! Earthquake.
- 3. A loss-of-coolant accident requiring actuation of the
- engineered safeguards.
' 4. mainJteam line nr feedwlter line breik. ve ADb ^ - - " --- - - -
Except that the inservice inspection due no later then October 20, g
{
1996, may be deferred until the next refueling outage (RFO 13), but j no later then4Juse 20,1997, whichever is earlier. !
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, MILLSTONE - . 2
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y /43 T-7 - KmendmentNo.22,37,ST,lff i , o.2 e OcMer .
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,'O April 9. 1986 TABLE 4.4-6 h~
9 4 STEAM GENERATOR TUBE INSPECTION 3RD SAW LE INSPECTION 9$ IST SAWLE INSPECTION 2ND SAMPLE INSPECTION Result Action' Required Sample Sire Result Action Required Result Action Required E
C-1 None N/A N/A N/A N/A F q 4 minimum of S tubes per N/A
.I
" C-2 Repair defective C-1 None N/A S.G.
tubes and i et C-2 Repair defective C-1 None additional l tubes and i t tubes in thi S.G.* additional tubes C-2 Repair defective in this S.G. tubes
- I Q 9 C-3 Perform action for C-3 result of first sample R.
S
. C-3 Perform action for 4 C-3 result of first N/A N/A C-3 Inspect all-
[ bes All other sample .
l N/A
- in this S. . repair S.G.s are None N/A l
! defectiv s and C-1 inspect tubes in Some S.G.s Perform action for l each other S.G.* C-2 but no C-2 result of second N/A N/A
~ k= additional sample -
- I Prompt notification S.G. are C-3 to NRC pursuant to Additional Inspect all tube's in a 10 CFR 50.72 S.G. is each S.G. and repair -
]
z P C-3 defective tubes.* N/A N/A
=
Prompt notification ,
to NRC pursuant to '
,M .
.A N U S=3 1 Where N is the number of steam generators in the unit, and n is the number of steam generators -
inspected
. n during an inspection e
-
- Repair of defective tubes shall be limited to plugging with the exception of those tubes which may be sleeved.
O Tubes with defective sleeves shall be plugged. ..
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2 -2_ 6 -- 9 (
'h' REACTOR COOLANT SYSTEM ary 15,1995 i SASE5 l only intended to pennit operation of the plant for a limited period of time not to exceed the next refueling outage so that maintenance can be perfonned on the block valve to eliminate the seat leakage condition or other similar concern.
The block valve should normally be available to allow PORV operation for auto-matic mitigation of overpressure events. The block valves should be returned to OPERABLE status prior to entering MODE 4 after a refueling outage.
If more than one PORV is inoperable and not capable of being manually cycled, it I is necessary to either restore at least one valve within the completion time of I hour or isolate the flow path by closing and removing the power to the associ-ated block valve, cooldown, depressurize, and vent the RCS.
I 3/4.4.4 PRESSURIZER i l
An OPERABLE pressurizer provides pressure control for the reactor coolant l system during operations with both forced reactor coolant flow and with natural l circulation flow. The minimum water level in the pressurizer assures the l pressurizer heaters, which are required to achieve and maintain pressure f control, remain covered with water to prevent failure, which occurs if the
! heaters are energized uncovered. The maximum water level in the pressurizer l
ensures that this parameter is maintained within the envelope of operation j assumed in the safety analysis. The maximum water level also ensures that the 1 ':
RCS is not a hydraulically sol'd system and that a steam bubble will be pro-
! vided to accomodate pressure surges during operation. The steam bubble also i protects the pressurizer code safety valves and power operated relief valve l against water relief. The requirement that a minimum number of pressurizer
! heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish and maintain natural circulation.
The requirement that 130 kW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of off-site power condition to maintain natural circulation at HOT STANDBY.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is IHSERT & & ;
(
MILLSTONE-UNIT 2 B 3/4 4-2a AmendmentNo.JJ,77,JJ,JJ,77,///
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hruary15,199 ItEACTOR C00LANT SYSTEN BASES f s _ _ - _ v v_ _-L --
evidence of mechanical damage or progressive degradation due to design, canufacturing errors, or inservice concitions 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.
g' , g _ _ _g The plant is expected to be operated in a manner such that the secondary WsE1tf i coolant will be maintained within those chemistryIflimits found to result in g the secondary coolant negligible corrosion of the steam generator tubes. .
chemistry is not maintained within these limits, localized corrosion may likely I result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 0.10 GPM, per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary leakage of 0.10 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and
- plugged.
Wastage-type defects are unlikely with proper chemistry 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 sleeving will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness, Sleeving repair will be limited to those steam generator tubes with a defect between the tube sheet and the first eggerate support. Tubes containing sleeves with imperfections exceeding the plugging limit will be plugged. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be immediately reported to the Connission pursuant to 10 CFR 50.72. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
l NILLSTONE - UNIT 2 8 3/4 4-2b Amendment No. JJ U , N , M , M , l sm in.1H,1H. AME l Odaq
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- 2. - 2 6 -9 f
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i INSERT B The inservice inspection of the steam generator tubes that was due no later than October 20,1996, may be deferred on a one-time-only basis by up to G-eiglet months (an entenoien-e6-46% beyond the maximum surveillance interval of 24 months) based on the following:
- The replacement steam generators have only been ir s/ ervice for one op(eratina ekOb cycle. \
- Forl10 months of the 3hmontMnterval between inspections, the i plant was shutdown and the steam generators were not exposed to the normal op,erating environment. I I - The tubes in the replacement steam generator are made of thermally-
! treated Inconel 690 which has been demonstrated to be more corrosion resistant than the material used in the original steam i generators. .
J i
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e i
e Docket No. 50-336 B15399 1
i i
4 Attachment 4 4
Millstone Nu:tlear Power Station, Unit No. 2 2
Proposed Revision to Technical Specifications Steam Generator Surveillance Requirement Extension 3
Retyped Version of Current Technical Specifications i
4 i
i i
4 1
October 1995 i
9'
REACTOR C00LANT SYSTEM SURVEILLANCE REQUIRENENTS (Continued) 4.4.5.1.3 Insoection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.* If two l consecutive ins)ections following service under AVT conditions, not including tie preservice inspection, result in all inspec-tion results falling into the C-1 category or if two consecutive inspections aemonstrate 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 4.4-6 at 40 month intervals fall into 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 inspec-tions satisfy the criteria of Specification 4.4.5.1.3.a; the interval may then be extended to a maximum of once per 40 months.
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-6 during the shutdown subsequent ,
to any of the following conditions: )
- 1. Primary-to-secondary tube leaks (not including leaks l l originating from tube-to-tube sheet welds) in excess j of the limits of Specification 3.4.6.2.. )
- 2. A seismic occurrence greater than the Operating Basis Earthquake. l l
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards. !
- 4. A main steam line or feedwater line break.
1 l
- Except that the inservice inspection due no later than October 20, 1996s maj be deferred until the next refueling outage (RF013), but no later than October 20, 1997, whichever is earlier, gI,L,LSTONE-UNIT 2 3/4 4-7 Amendment No. 77,77,77,Jpg, l
. ~ .
TABLE 4.4-6 g5
. r-STEAN GENERATOR TUBE INSPECTION u; IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION
~$ Sample Size Result Action Required Result Action Required Result Action El Reautred i A minimum of C-1 None N/A N/A N/A N/A l S tubes per C-2 Repair defective C-1 None N/A N/A
- h. S.G. tubes and inspect
-4 additional 2S tubes n> in this S.G.*
l C-2 Repair defective C-1 None tuMs and inspect additional 45 tubes C-2 Repair i in this S.G.* defective tubes
- C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A C-3 result of first u, sample
]g C-3 Inspection all tubes All other None N/A N/A in this S.G., repair S.G.s are i' defective tubes and C-1
- ( inspect 2S tubes in each other S.G.*
Some S.G.s Perform action for N/A N/A Prompt notification C-2 but no C-2 result of second to NRC pursuant to additional sample i !I 10 CFR 50.72 S.G. are C-3 '
$ Additional Inspect all tubes in N/A N/A Et S.G. is C-3 each S.G. and re
@ defective tubes.* pair ,
r* Prompt notification ne to NRC pursuant to P 10 CFR 50.72
~~
0$ N S= 3 % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected g3 n during an inspection 3,
- Repair of defective tubes shall be limited to plugging with the exception of those tubes which may be sleeved.
Ia Tubes with defective sleeves shall be plugged.
D.
- - ~. . _ ..- .. . . . - - . - - . - ..-. - - . . . ..
t
)* REACTOR COOLANT SYSTEM BASES i i
i l only intended to permit operation of the plant for a limited period of time not !
- to exceed the next refueling outage so that maintenance can be performed on the
- block valve to eliminate the seat leakage condition or other similar concern. i
! The block valve should normally be available to allow PORV o)eration for auto- ;
matic mitigation of overpressure events. The block valves s1ould be returned to i
OPERABLE status prior to entering MODE 4 after a refueling outage. i i If more than one PORV is inoperable and not capable of being manually cycled, it !
! is necessary to either restore at least one valve within the completion time of i i hour or isolate the flow path by closing and removing the power to the associ-
!- ated block valve, cooldown, depressurize, and vent the RCS.
i 3/4.4.4 PREESURIZER .
! An OPERABLE pressurizer provides pressure control for the reactor coolant j system during operations with both forced reactor coolant flow and with natural
- circulation flow. The minimum water level in the pressurizer assures the
{ pressurizer heaters, which are required to achieve and maintain pressure i i control, remain covered with water to prevent failure, which occurs if the r
- heaters are energized uncovered. The maximum water level in the pressurizer l
- ensures that this parameter is maintained within the envelope of operation ;
- assumed in.the safety analysis. The maximum water level also ensures that the '
j RCS is not a hydraulically solid system and that a steam bubble will be pro-
- vided to accommodate pressure surges during operation. The steam bubble also l protects the pressurizer code safety valves and power operated relief valve against water relief. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor 1 Coolant System pressure and establish and maintain natural circulation.
l l The requirement that 130 kW of pressurizer heaters and their associated
- controls be capable of being supplied electrical power from an emergency bus
- provides assurance that these heaters can be energized during a loss of off- ,
- site power condition to maintain natural circulation at H0T STANDBY.
- '
i l 3/4.4.5 STEAM GENERATORS i
n The Surveillance Requirements for inspection of the steam generator
! tubes ensure that the structural integrity of this portion of the RCS will be i
- 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 j 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 '
i taken.
i i
(
! NILLSTONE-UNIT 2 B3/44-2a Amendment No. 77, 77, pg, pp, 77,
- 177, ,
l I
REACTOR COOLANT SYSTEN ,
BASES The inservice inspection of the steam generator tubes that was due no later l than October 20, 1996, may be deferred on a one-time-only basis by up to 12 months '
(beyond the maximum surveillance interval of 24 months) based on the following:
l
- The replacement steam generators have only been in service for one operating l cycle.
- For at least 10 months of the 36-mo.m. interval between inspections, the plant I was shut down and the steam generators were not exposed to the normal operating environment. !
i
- The tubes in the replacement steam generator are made of thermally-treated Inconel 690 which has been demonstrated to be more corrosion resistant than the material used in the original steam generators.
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 l chemistry is not maintained within these limits, localized corrosion may likely i result in stress corrosion cracking. 1 The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 0.10 GPM, per steam generater). Cracks having a primary-to-secondary leakage less than l this limit during operation will have an adequate margin of safety to withstand ,
the loads imposed during normal operation and by postulated accidents. )
Operating plants have demonstrated that primary-to-secondary leakage of 0.10 ,
gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry 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 sleeving will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Sleeving repair ,
will be limited to those steam generator tubes with a defect between the tube '
sheet and the first eggerate support. Tubes containing sleeves with imperfections exceeding the plugging limit will be plugged. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. .
Whenever the results of any steam generator tubing inservice inspection i fall into Category C-3, these results will be immediately reported to the Commission pursuant to 10 CFR 50.72. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
i l NILLSTONE - UNIT 2 B 3/4 4-2b Amendment No. 77, 77, J7, 77, 77, om yyy,Jgy,177,17),