ML042920289

TS for Beaver Valley Power Station, Unit No. 1 (BVPS-1)

ML042920289
Person / Time
Site:	Beaver Valley
Issue date:	10/15/2004
From:	NRC/NRR/DLPM
To:
References
TAC MC3671
Download: ML042920289 (12)

Text

Table Index (cont.)

TABLE TITLE PAGE 3.3-11 4.3-7 4.4-1 4.4-2 4.4-3 4.4-12 3.7-1 3.7-2 4.7-2 3.8-1 3.9-1 Accident Monitoring Instrumentation Accident Monitoring Instrumentation Surveillance Requirements Minimum Number of Steam Generators to be Inspected During Inservice Inspection Steam Generator Tube Inspection Reactor Coolant System Pressure Isolation Valves Primary Coolant Specific Activity Sample and Analysis Program OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Steam Line Safety Valves Per Loop Secondary Coolant System Specific Activity Sample and Analysis Program Battery Surveillance Requirements Beaver Valley Fuel Assembly Minimum Burnup vs. Initial U235 Enrichment For Storage in Region 2 Spent Fuel Racks 3/4 3/4 3-51 3-52 3/4 4-lOh I

I 3/4 3/4 4-10i 4-14b 3/4 4-20 3/4 7-2 3/4 3/4 3/4 3/4 7-4 7-9 8-9a 9-15 BEAVER VALLEY -

UNIT 1 XVII (Next page is XIX)

Amendment No.

2.62

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

2.

Tubes in those areas where experience has indicated potential problems, and

3.

Except for Alloy 800 leak limiting sleeves, at least 3 percent of the total number of sleeved tubes in all three steam generators.

A sample size less than 3 percent is acceptable provided all the sleeved tubes in the steam generator(s) examined during the refueling outage are inspected.

All inservice Alloy 800 leak limiting sleeves shall be inspected over the full length using a plus point coil or equivalent qualified technique during each refueling outage.

These inspections will include both the tube and the sleeve, and

4.

A tube inspection pursuant to Specification 4.4.5.4.a.8.

If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

5.

Indications left in service as a result of application of the tube support plate voltage-based repair criteria (4.4.5.4.a.10) shall be inspected by bobbin coil probe during all future refueling outages.

c.

The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1.

The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and

2.

The inspections include those portions of the tubes where imperfections were previously found.

d.

Implementation of the steam generator tube-to-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.

BEAVER VALLEY - UNIT 1 3 /4 4-9 Amendment No. 262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) e.(3) Implementation of the steam generator WEXTEX expanded region inspection methodology (W*), requires a 100 percent rotating probe inspection of the hot leg tubesheet W*

distance.

The results of each sample inspection shall be classified into one of the following three categories:

Catecrory Inspection Results C-1 Less than 5 percent 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 percent of the total tubes inspected are defective, or between 5

percent and 10 percent of the total tubes inspected are degraded tubes.

C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective.

Note:

In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10 percent) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Inspection Frecuencies -

The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a.

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under All Volatile Treatment (AVT) conditions, not including the preservice inspection, result in all inspection 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 months.

(3) Applicable only to Cycle 17.

BEAVER VALLEY - UNIT 1 3/4 4-10 Amendment No. 262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b.

If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results 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 a subsequent inspection demonstrates that a third sample inspection is not required.

c.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1.

Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,

2.

A seismic occurrence greater than the Operating Basis Earthquake,

3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or

4.

A main steamline or feedwater line break.

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 drawings or specifications.

Eddy-current testing indications below 20 percent of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

Degradation means a

service-induced

cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.

3.

Degraded Tube means a tube or sleeve containing imperfections greater than or equal to 20 percent of the nominal wall thickness caused by degradation.

4.

Percent Degradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.

BEAVER VALLEY - UNIT 1 3/4 4-10a Amendment No-267

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

5.

Defect means an imperfection of such severity that it exceeds the plugging or repair limit.

A tube containing a defect is defective.

Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6.

Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection.

The plugging or repair limit imperfection depths are specified in percentage of nominal wall thickness as follows:

a)

Original tube wall 40%

1.0) This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.

Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.

2.0)(3) This definition does not apply to service induced degradation identified in the W*

distance.

Service induced degradation identified in the W* distance or less than eight inches below the top-of-tube sheet (TTS),

which ever is greater, shall be repaired on detection.

b)

ABB Combustion Engineering TIG welded sleeve wall 32%

c)

Westinghouse laser welded sleeve wall 25%

d)

Westinghouse Alloy 800 leak limiting sleeve(3):

Plug on detection of any service induced imperfection, degradation or defect in the (a) sleeve and/or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steamline or feedwater line break as specified in 4.4.5.3.c, above.

(3) Applicable only to Cycle 17.

BEAVER VALLEY -

UNIT 1 3/4 4-10b Amendment No. 262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

8.

Tube Inspection means an inspection of the steam generator tube from the point of entry (hot-leg side) completely around the U-bend to the top support of the cold-leg, excluding the portion of the tube within the tubesheet below the W* distance, the tube to tubesheet weld and the tube end extension.

This exclusion is applicable only to Cycle 17.

This exclusion does not apply to steam generator tubes with sleeves installed within the tubesheet region.

9.

Tube Repair refers to sleeving which is used to maintain a tube in-service or return a tube to service.

This includes the removal of plugs that were installed as a corrective or preventive measure.

The following sleeve designs have been found acceptable:

a)

ABB Combustion Engineering TIG Welded Sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.

b)

Westinghouse laser welded sleeves, WCAP-13483, Revision 1.

c)

Westinghouse Alloy 800 leak limiting sleeves, WCAP-15919-P, Revision 00.(

10.

Tube Support Plate PlucrqinQ Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates.

At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:

a)

Steam generator

tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin 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 4.4.5.4.a.10.c below.

(3) Applicable only to Cycle 17.

BEAVER VALLEY - UNIT 1 3/4 4-10c Amendment No - 2.62

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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 ttp n or equal to the upper voltage repair limit' may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

Steam generator

tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltpe greater than the upper voltage repair limit will be plugged or repaired.

d) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.

The mid-cycle repair limits are determined from the following equations:

VMM =

VAL 1.0 + NDE + Gr (CL -At)

CL VLRL = VMURL - (VUL

- VLa)( CL')

CL where:

VuL

=

upper voltage repair limit VLRL

=

lower voltage repair limit VxuL

=

mid-cycle upper voltage repair limit based on time into cycle VmLRL

=

mid-cycle lower voltage repair limit based on V1 RL and time into cycle (1) The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

BEAVER VALLEY - UNIT 1 3/4 4-10d Amendment No. 262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

At

=

length of time since last scheduled inspection during which Van and VLpL were implemented CL

=

cycle length (the time between two scheduled steam generator inspections)

VSL

=

structural limit voltage Gr

=

average growth rate per cycle length NDE

=

95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC) (2)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.

11. (3) a)

Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet, as determined by eddy current testing.

b)

W* Distance is the non-degraded distance from the top of the tubesheet to the bottom of the W*

length including the distance from the top of the tubesheet to the bottom of the WEXTEX transition (BWT) and Non-Destructive Examination (NDE) measurement uncertainties (i.e.,

W*

distance =

W* length + distance to BWT + NDE uncertainties).

c)

W* Lenath is the length of tubing below the bottom of the WEXTEX transition (BWT) which must be demonstrated to be non-degraded in order for the tube to maintain structural and leakage integrity.

For the hot leg, the W* length is 7.0 inches which represents the most conservative hot leg length defined in WCAP-14797, Revision 2.

(2) The NDE is the value provided by the NRC in GL 95-05 as supplemented.

(3) Applicable only to Cycle 17.

BEAVER VALLEY - UNIT 1 3/4 4-10e Amendment No.

262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit) required by Table 4.4-2.

4.4.5.5 Reports

a.

Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be submitted in a Special Report in accordance with 10 CFR 50.4.

b.

The complete results of the steam generator tube and sleeve inservice inspection shall be submitted in a Special Report in accordance with 10 CFR 50.4 within 12 months following the completion of the inspection.

This Special Report shall include:

1.

Number and extent of tubes and sleeves inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.

Identification of tubes plugged or repaired.

c.

Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation.

The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

d.

For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise:

1.

If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

For Cycle 17, the postulated leakage resulting from the implementation of the voltage-based repair criteria to tube support plate intersections shall be combined with the postulated leakage resulting from the implementation of the W* criteria to tubesheet inspection depth.

BEAVER VALLEY -

UNIT 1 3/4 4-10f Amendment No-262

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

2.

If circumferential crack-like indications detected at the tube support plate intersections.

are

3.

If indications are identified that extend beyond the confines of the tube support plate.

4.

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

5.

If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured2 end-of-cycle) voltage distribution exceeds 1 X 10, notify the Commission and provide an assessment of the safety significance of the occurrence.

e. (3) The aggregate calculated steam line break leakage from the application of tube support plate alternate repair criteria and W* inspection methodology shall be submitted in a Special Report in accordance with 10 CFR 50.4 within 90 days following return of the steam generators to service (MODE 4).

In addition, the total number of indications that are identified from 1R16 rotating probe inspections that are performed as part of the W*

inspections will be included in this report.

(3) Applicable only to Cycle 17.

I BEAVER VALLEY - UNIT 1 3/4 4-10g Amendment No.262

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Three Three First Inservice Inspection All Two Second & Subsequent Inservice Inspections One (1)

One (2)

Table Notation:

(1) The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 9 percent of the tubes if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.

Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

(2) The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in (1) above.

BEAVER VALLEY -

UNIT 1 3/4 4-10h Amendment No.2A2

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Result Action Required Result Action Required Result Action Required Size A minimum C-1 None N/A N/A N/A N/A of S tubes C-2 Plug or repair C-1 None N/A N/A per S.G.

defective tubes and C-2 Plug or repair defective C-1 None inspect additional 2S tubes and inspect additional C-2 Plug or repair tubes in this S.G.

4S tubes in this S.G.

l defective tubes C-3 Perform action for C-3 result l_

of first sample C-3 Perform action for C-3 N/A N/A result of first sample l

C-3 Inspect all tubes in All other None N/A N/A this S.G., plug or S.G.s are repair defective tubes C-1 and inspect 2S tubes in each other S.G.

l Some S.G.s Perform action for C-2 N/A N/A Notification to NRC C-2 but no result of second sample pursuant to additional Specification 6.6 S.G.s are C-3 Additional Inspect all tubes in each N/A N/A S.G. is S.G. and plug or repair C-3 defective tubes.

Notification to NRC pursuant to Specification 6.6.

s = 9 %

Where n is the number of steam generators inspected during an inspection.

n BEAVER VALLEY -

UNIT 1 3/4 4-10i Amendment No. 262 I