ML20212F047

Proposed Tech Specs Pages Revising TS Section 4.12,allowing Use of voltage-based SG Tube Repair Criteria

ML20212F047
Person / Time
Site:	Prairie Island
Issue date:	10/28/1997
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20212F045	List: ML20212F041 ML20212F047
References
NUDOCS 9711040154
Download: ML20212F047 (13)

Text

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1. 4 Attachment Revised Technical Specification Pages l

1 l

9711040154 971028 PDR ADOCK 05000282 P pm

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.TS vi

-REV'-  ;

l TABLE OF CONTENTS (Continued) <

TS SECTION^ TITLE PACE

~4.12 Steam Generator Tube Surveillance' TS.4.12-1  !

A'. Steam Generator Sample Selection and. TS.4.12-1 Inspection-B.: Steam Generator Tube Sample Selection- TS.4.12 1

= and_-Inspection -

C.: Inspection Frequencies TS.4.12 3  ;

D. Acceptance; Criteria'. TS.4.12-4  !

E. Reports 'TS.4.12-7 l 4.13 . Snubbers _ _

TS.4.13 1 4.14 . Control-Room Air Treatment System Tests TS.4.14-1 4'15. Spent Fuel Pool Special-Ventilation _ System TS.4.15-l'.

.4.161 Deleted 4.17- Deleted' 4.18 Reactor _ Coolant Vent System Paths-- TS.4.18-1 A. Vent Path operability TS.4.18-1

'B. System Flow Testing _ __

TS.4.18-1

'4.19- Auxiliary Building Crane Lifting Devices- TS.4.19-1 4~20 Spent. Fuel' Pool Storage Configuration TS.4.20-1 L

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TS.3.1-9'

! REV-p

-3;1.C.2' e. If the total reactor coolant-system to secondary coolant system-

. leakage 1through any.one steam generator.of a_ unit exceeds 150:

gallons ~per day (GPD)' within one hour initiate action to place

' the unit.in HOT SHUTDOW and be in at least HOT SHUTDOW within

. the next 6 hturs and be in COLD SHUTDOW within the following 30 -

hours and periorm an inservice steam generator tube inspection in accordance_with. Technical Specification 4.12.

3. Pressure' Isolation Valve Leakare Leakage through the pressure isolation valves shall not exceed the maximum allowable leakage specified in Specification 4.3 when i reactor coolant synem average temperature exceeds 200'F. If the maximum allowable leakage is exceeded, within one hour initiate the action necessary to place tne unit in HOT SHUTDOW, and be -

in at least HOT SHUTDOW 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

- SHUTDOW 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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. 4 TS.4.12 3

. REV

5. Indications left in service as a result of application of tube support plate voltage based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

6. Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest cold leg tube support plate 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.

C. Inspection Freauencies The above required in-service inspections of steam generator tubes shall be performed at the following frequencies:

1. In-service inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C 1 category or if two consecutive inspections l 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.

2. If the results of the inservice inspection of a steam generator conducted in accordance with Table TS.4.12-1 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 4.12.C.1; the interval may then be extended to e maximum of once per 40 months.

3. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table TS.4.12-1 during the shutdown subsequent to any of the following conditions.

(a) Primary-to secondary tube leaks (not including leaks originating from tube to tube sheet welds) in excess of the limits of Specification 3.1.C.6.

(b) A seismic occurrence Breater than the Operating Basis Earthquake.

(c) A loss of-coolant accident requiring actuation of the engineered safeguards.

(d) A main steam line or feedwater line break.

- TS,4.12-4 REV D. Acceptance Criteria

1. .As used in this Specificatioa:

(a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

(b) Dent adation means a service induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

(c) Denraded Tube means a tube containing imperfections h20% of the nominal wall thickness caused by degradation.

(d)  % Derradation means the percentage of the tube wall thickness affected or removed by degradation.

(e) Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(f) Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection and is equal to 50% of the nominal tube wall thickness. If significant general tube thinning occurs, this criteria will be reduced to 40% wall penetration. This definition does not apply to the portion of the tube in the tubesheet below the F* distance provided the tube is not degraded (i.e. , no indications of cracks) within the F*

distance for F* tubes. The repair limit for the pressure boundary region of any sleeve is 31% of the nominal sleeve wall thickness. This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to Specification 4.12.D.4 for the repair limit appliccble to these intersections.

(g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a steam line or feedwater line break.

(h) 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.

(i) Sleeving is the repair of degraded tube regions using a new

. Alloy 690 tubing sleeve inserted inside the parent tube and sealed at each end by welding or by replacing the lower weld in a full depth tubecheet sleeve with a hard rolled joint. The new sleeve becomes the pressure boundary spanning the original degraded tube region.

- . TS.4.12-5 )

REV I (j ) F* Distanci is the distance from the bottom of the hardroll transition toward the bottom of the.tubesheet that has been conservatively determined to be 1.07 inches (not including eddy cerrent uncertainty).

(k) F* Tube is a tube with degradation, below the F* distance, equal to or Breater than 404, and not degraded (i.e. , no indications of cracking) within the F* distance.

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the repair ifmit and all tubes containing through wall cracks or classify as Pa . tubes) required by Tables TS.4.12-1 and TS.4.12 2,

3. Tube rupair, after October 1, 1997, using Combustion Engineering volded sleeves shall be in accordance with the methods described in the following:

CEN-629-P, Revision 2 " Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves";

CEN-629-P, Addendum 1, Revision 1, " Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Siceves"

4. Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates, At tube support plate intersections, the repair limit is based on maintaining steam generator 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 voltages less than or equal to 2.0 volts will be allowed to remain in service.

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 2.0 volts, will be repaired or plugged, except as noted in Specification 4.12.D.4.c below.

c. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit, may remain in service if a rotating pancake coil (or comparable examination technique) inspection does not detect degradation. Steam generator tubes, with indications of outside dfameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair' limit will be plugged or repaired.

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Ts.4.13 6- l

.REV' ,

[d.' If/ an unscheduled mid-cycle inspection is perfoined, the following.  ;

mid cycle repair' limits apply instead of the limits in '

. Specifications.4.12.D.4.a. b and'c. The mid cycle repair limits Care determined : from the -i'ollowing -equations:

1.0+NDE+G

}AC Va = V. n-(Vm-2.0)( I,}A g c);

whete:

Vuiu, - upper voltage repair limit v uu. - lower voltage repair limic V m - mid cycle upper voltage repair limit based on time into cycle V m - mid cycle lower voltage repair limit based on V m and time into' cycle 4

At - length of time since last scheduled inspection during which Vuiu. and Vuu. were implemented CL - cycle length (time between two scheduled steam generator inspections)

Vst - structural limit voltage Cr - average growth rate per cycle length ,

NDE - 95 percent cumulative probability allowance for

- nandestructive examination uncertainty (i.e. , a value of 20 percent has been! approved by the NRC)

- Implementation of these mid cycle repair limits should follow.the same approach as described in Specifications 4.12.D.4.a b and c.

Note:-The upper voltage repair limit is_ calculated according to the methodology in Generic Letter 95-05.as' supplemented.

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• TS.4.12 7

. REV E. - Reports

1. Following each_in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes  ;

plugged or sleeved in each steam generator shall be reported to the conmission within 15 days.

2. The results of steam generator tube inservice inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generatot tube inservice inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection. These reports shall include: (1) number and extent of ,

tubes inspected, (2) location and percent of wall thickness penetration for each indication of an imperfection and (3) identification of tubes plugged or sleeved. ,

3. Results of stea.a generator tube inspections which fall into Category C 3 requira notification to the Commission prior to resumption of plant opnration, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures caken to prevent recurrence.

4. The results of inspections performed under Specification 4.12.B for all tubes that have defects below the P* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:

a. Ident.fication of F* tubes, and

b. Location and extent of degradation.

5. For implementation of the voltage based repair criteria to tube support

, plate intersections, notify the NRC staff prior to returning the steana generators to service should any of the following conditions ariae:

a. 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 steamline break) for the next operating cycle.

'b. If circumferential crack like indications are detected at the tube support plate intersections.

c. ~1f indications'are identified that extend beyond the confines of the tube support plate,

d. If indications are identified at the tube support plate elevations that are attributabla to primary water stress corrosion cracking.

e. 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 distribution exceeds 1 x 10 a, notify the NRC and provide an assessment of the safety significance of the occurrence.

7 B.3.1 7 [

' REV f

341 REACTOR C001 ANT SYSTM r

Bases continued ,

C. Reactor Coolant System Leaksgo Leakage from the reactor coolant system is collected in the containment  !

or_by other systems. These systems are the main steam system, conden. 1 sate and feedwater system and the chemical and volume control system. L i'

Detection of leaks from the reactor coolant system is by one or more of the following (Referenct 1): i i

1. An increased amount of makeup water required to maintain normal [

' level in the pressurizer.

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2. A high temperature alarm in the leakoff piping provided to collect ,

reactor head flange leakage. l t

3. Containment sump water level indication, ,

4. - Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the reactor  !

coolant, the radiation monitoring system provides a sensitive indica-tion of primary system leakage. Radiation monitoro which indicate  ;

primary system leakage include the containment air particulate and gas +

monitors, the area radiation monitors, the condenser air ejector [

monitor, the component cooling water monitor, and the steam generator blowdown monitor (Reference 2). .

The historical leak rate limit of 1 gpm corresponded to a through wall crack less than 0.6 inches long based on test data. Steen generator tubes having_a 0,6 inch long through wall crack have been shown to resist  ;

failure at pressures resulting from normal operation, LOCA, or steam line break accidents (Reference 3).  ;

The leakage limits incorporated into Specification 3.1.C for [

implementation of the Steam Generator Voltage Based Alternate Repair

• Criteria are more restrictive than the standard operating leakage limits and are intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extend ,

outside the thickness of the tube support plate. Hence, the reduced leakage limit, when combined with an effective leak rate monitoring- .

program, provides additional assurance that should a significant leak be i experienced in service, it wi'l be detected, and the plant shut down in a '

timely manner, Specification 3.1.C.3 specifies actions to be taken in the event of

-failure or excessive leakage _of a check valve which isolates the high pressure reactor coolant system from the low pressure RHR system niping, v

References

'1 _USAR, Section 6.5

2. USAR, Section 7.5.1' ~

3. LTestimony by J Knight in the Prairie Island public hearing on

-January 28,:1975,;pp 13 17.

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B.4.12 1 REV 4.12 STEAM GENERATOR TUBE SURVEILIANCE Bases 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 r is based on a modification of Regulatory Guide 1.83, Revision 1. '

In service 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 in service conditions that lead to corrosion. In service 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. I At the request of the NRC, a requirement for in service inspection of at least 20% of the total number of sleevas in service in both steam generators was added to TS 4.12.B. In addition. Table TS 4.12 2 was added to provide sample expansion requirements based on the results of the initial sample inspection similar to Table 4.12 1 This type of sample size and expansion requirement is consistent with the EPRI PWR Steam Generator Examination Guidelines. The sample selection is applied to each type of =leeve. Types of sleeves are categorized by such characteristics as the installation vendor, the sleeve material, the type of joint such as lower edge' weld or lower hard roll joints, the sleeve location such as tube support plate or tubesheet and whether not the welded joints have 3 received post weld heat treatment.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameters found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameters, localized corrosion would most likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator leakage between the primcry coolant system and the secondary coolant system (primary to secondary leakage - 150 gallons per day from one steam generator). Historically, cracks not addressed by voltage based alternate repair criteria and having a primary to secondary leakage less than 1.0 gpm (1440 gallons per day) . ,

during operation would have an adequate margin of safety against Tsilure '

due to loads imposed by design basis accidents (Reference 1). Operational experience has demonstrated that primary to secondary leakage as low as 5 gallons per day will be detected by secondary system radiation monitors.

-To provide defense in depth for the voltage based repair criteria, leakage in excess of 150 gallons per day from one steam generator will require plant shutdown and an unscheduled eddy current inspection, during which the leaking tubes will be located and plugged or sleeved.

Vastage type defects are unlikely with proper chemistry treatment of secondary coolant. However, even if this type of defect occurs it will be found during scheduled in service steam generator tube inspections, Repair or plugging will be required of all tubes with imperfections that could develop defects having less than the minimum acceptable wall thickness prior to' the next inservice inspection which, by the definition of Specification 4.12.D.1.(f), is 50% of the tube or sleeve nominal wall thickness. Wastage type defects having a wall thickness greater than l

B.4.13 3 '

. REV 4.12 STEAM CENERATOR TUBE SURVEILIANCE Bauen continued 0.025 inches will have adequate margins of safety against failure due to

• loads imposed by normal plant operation and design basis accidents (Reference 1). Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage typ) defects that have penetrated 20% of the original 0.050 inch wall thickness (Reference 2). l Plugging or sleevin5 is not required for tubes meeting the F* critoria. >

The F* distance will be controlled by a combination of eddy current inspection and/or process control. For a new additional roll expansion, the requirement will be at least 1.2 inches of new hard roll. This is controlled by the length of the rollers (1.25 inch effectiva length). The distance from the original roll transition zone is also controlled by the process in that the lower end of the new roll expansion is located one i inch above the original roll expansion. In the case of the new roll, eddy current examination will confirm there are no indications in the new roll region and that there is a new roll region with well defined upper and lower expansion transitions.

When eddy current exar! nation, alone, must determine the F* distance, such as in the existing hard roll region, or when multiple lenSths of additional hard rolls have been added, the eddy current uncertainty is qualified by testing against known standards. That value is expected to be 0.18 inches. Therefore, the F* distance measured by eddy current (sum of 1.07 and 0.18) will be conservatively set at 1.3 inches.

When more than one Alternate Repair Criteria are used, the summat'on of leakage from all tubes left in service by all repair criteria must be less than the allowable leakage for the most limiting of those Alternate Repair Criteria.

Whenever the results of any steam generator tubing in service inspection fall into Category C 3, these results will be promptly reported to the Commission prior to resumption of plant operation. Such cases will be considered by the Commission on a ase by. case basis and may result in a requirement for analysis, laboratory examinations, teste, additional eddy current inspection, and revision of the Technical Specifications, if necessary.

Degraded stum generator tubes may be repaired by the installation of sleeves uhich span the section of de5raded steam generator tubing. A steam generator tube with a sleeve installed meets the structural requirements of tubes whien are not degraded.

The following sleeve deuigns have been found acceptable by the NRC Staff: *

a. Westinghouse Mechanical Sleeves (WCAP 10757)

b. Westinghouse Brazed Sleeves (WCAP 10820)

c. Combustion Engineering Leak Tight Sleeves (CEN 294 P, for sleeves installed prior to October 1, 1997)

d. Combustion Engineering Leak Tight Sleeves (CEN 629 P, for sleeves installed after 0ctober 1, 1997)

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B.4.12-3

- REV 4.12 STEAM CENERATOPXQURVEILIANCE pases continued Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval prior to their use in the repair of degraded steam generator tubes. The submittals related to other sleeve designs shall be made at least 90 days prior to use.-

Tube Sungort Plate Repair Limit i

The voltage. based repair limits of Specification 4.12.D.4 implement the guidance in Generic Letter 95 05 and are applicable only to Westinghouse.

designed steam generators with outsido 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 steam generator tube degradation nor are they appiicable to ODSCC that occurs at other locations within the steam generator. 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 plats. Refer to Generic Letter 95 05 for additional description of the degradation morphology.

Implementation of Specification 4.12.D.4 requires a derivation of the voltage structural limit from the-burst versus voltage empirical correlation and then the subsequent derivation of the vc.itage repair limit from the structural limit (which is then implemented by this surveillance).

The voltago 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 raaterial properties at 650'F (i.e. , the 95 percent LTL curve) . The vultage 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; Vma, is determined from the structural voltage limit by applying the following equation: ,

Vma. - Vst.

• Vor

• Vsor whern Va, represents the allowance for flaw growth between inspections and Vsos represents the allowance for potential sources of error in the measurernent of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limir are discussed in Generic letter 95 05.

The mid. cycle equation in Specification 4.12.D.4 should only be used during unplanned inspections in which eddy current data is acquired for indications at-the tube-support plates.

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B.4.12 4

- REV 4.12 STEAM CENERATOR TUBE SURVEII.!ANCE Bases continued specification 4.12.h.5 implements several reporting requirements recommended by Generic Letter 95 05 for situations which the NRC wants to be notified prior to returning the steam generators 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 I Ceneric Letter 95 05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions ,

prior to returning the steam Senerators to service. Note that if leakage ,

and conditional burst probability were calculated using the measured EOC  :

voltage distribution for the purposes of addressing Generic Lotter 95 05 Section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected E00 voltage distribution should be provided per Generic Letter.

95 05-Section 6.b (c) criteria.

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References

1. Testimony of J Knight in the Prairie Island Public Hearing- on 1/28/75-t

2. Testimony of L Frank in the Prairio Island Public Hearing on 1/28/73

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