ML070950265

Technical Specification Pages Steam Generator Tube Repair in the Tubesheet Amendment

ML070950265
Person / Time
Site:	Point Beach
Issue date:	04/04/2007
From:	Raghavan L NRC/NRR/ADRO/DORL/LPLIII-1
To:	Koehl D Nuclear Management Co
Milano P, NRR/DORL/LPLIII-1, 415-1457
Shared Package
ML070800705	List: ML070800625 ML070950265
References
TAC MD2583
Download: ML070950265 (11)

Text

D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NMC to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Levels NMC is authorized to operate the facility at reactor core power levels not in excess of 1540 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 226, are hereby incorporated in the renewed operating license.

NMC shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee 2 is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

2 Reference to the licensee in License Conditions 4.C, 4.E and 4.H refers to Wisconsin Electric Power Company and is maintained for historical purposes.

Renewed License No. DPR-24 Amendment No. 226

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.8 ELECTRICAL POW ER SYSTEMS ................................................................ 3.8.1-1 3.8,1 AC Sources-Operating .......................................................................... 3.8.1-1 3.8,2 AC Sources-Shutdown .......................................................................... 3.8.2-1 3.8,3 Diesel Fuel Oil and Starting Air ........................... 3.8.3-1 3.8,4 DC Sources-Operating .......................................................................... 3.8.4-1 3.8,5 DC Sources-Shutdown .......................................................................... 3.8.5-1 3.8,6 Battery Cell Parameters ........................................................................ 3.8.6-1 3.8.7 Inverters-Operating ................................................................................ 3.8.7-1 3.8&8 Inverters-Shutdown ............................................................................... 3.8.8-1 3.8&9 Distribution Systems-Operating ............................................................. 3.8.9-1 3.8,10 Distribution Systems-Shutdown ............................................................ 3.8.10-1 3.9 REFUELING OPERATIONS .......................................................................... 3.9.1-1 3.9,1 Boron Concentration .............................................................................. 3.9.1-1 3.9.2 Nuclear Instrumentation ........................................................................ 3.9.2-1 3.9,3 Containment Penetrations ..................................................................... 3.9.3-1 3.9A4 Residual Heat Removal (RHR) and Coolant Circulation-High W ater Level ....................................................... 3.9.4-1 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-Low W ater Level ......................................................... 3.9.5-1 3.9.6 Refueling Cavity W ater Level ................................................................ 3.9.6-1 4.0 DESIGN FEATURES .................................................................................... 4.0-1 4.1 Site Location .......................................................................................... 4.0-1 4.2 Reactor Core ......................................................................................... 4.0-1 4.3 Fuel Storage .......................................................................................... 4.0-2 5.0 ADMINISTRATIVE CONTROLS .................................................................... 5.1-1 5.1 Responsibility ........................................................................................ 5.1-1 5.2 Organization .......................................................................................... 5.2-1 5.3 Unit Staff Qualifications .............................. .......................................... 5.3-1 5.4 Procedures ............................................................................................ 5.4-1 5.5 Programs and Manuals ......................................................................... 5.5-1 5.6 Reporting Requirements ........................................................................ 5.6-1 5.7 High Radiation Area .............................................................................. 5.7-1 Point Beach iii Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Pro-gram (continued) for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 500 gallons per day per SG.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth-based criteria:

1. For Unit 1 Refueling Outage 30 and the subsequent operating cycle, flaws found in the portion of the tube below 17 inches from the top of the hot leg tubesheet do not require plugging. All tubes with flaws identified in the portion of the tube within the region from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged. This alternate tube repair criteria is not applicable to the tube at row 38 column 69 in the A steam generator, which is not expanded the full length of the tubesheet.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 1 Refueling Outage 30 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded when the alternate repair criteria in TS 5.5.8.c are implemented. This exclusion does not apply to the tube at row 38 column 69 in the A steam generator, which is not expanded the full length of the tubesheet. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

Point Beach 5.5-8 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 229

Programs and Manuals 5.5 5.5 Programs and Manuals

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. i. Unit 1 (alloy 600 Thermally Treated tubes): Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

ii. Unit 2 (alloy 690 Thermally Treated tubes): Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

Point Beach 5.5-9 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 229

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.10 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of the Control Room Emergency Filtration System (F-16) at the frequencies specified in Regulatory Guide 1.52, Revision 2, and in accordance with ASTM D3803-1989 and the methodology of ANSI N510-1980, as prescribed below.

a. Demonstrate for the Control Room Emergency Filtration System (F-1 6) that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass _<1.0% when tested in accordance with the methodology of ANSI N510-1980, Section 10, excluding subsection 10.3, at a system flowrate of 4950 cfm +/- 10%.

b. Demonstrate for the Control Room Emergency Filtration System (F-16) that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0% when tested in accordance with the methodology of ANSI N510-1980, Section 12, excluding subsection 12.3, at a system flowrate of 4950 cfm +/- 10%.

Point Beach 5.5-10 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

c. Demonstrate for the Control Room Emergency Filtration System (F-16) that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with the methodology of ANSI N510-1980, Section 13, excluding subsection 12.3, shows the methyl iodide penetration _<1.0%, when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and a relative humidity of 95%, applying the tolerances of ASTM D3803-1989.

d. Demonstrate for the Control Room Emergency Filtration System (F-16) that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than 6 inches of water when tested in accordance with the methodology of ANSI N510-1980, Sections 10 and 12, excluding subsections 10.3 and 12.3, at a system flowrate of 4950 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.11 Explosive Gas Monitorinc Program This program provides controls for potentially explosive gas mixtures contained in the on-service Gas Decay Tank.

The program shall include a limit for oxygen concentration in the on-service Gas Decay Tank and a surveillance program to ensure the limit is maintained. This limit shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion).

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies.

Point Beach 5.5-11 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Diesel Fuel Oil Testing Pro-gram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. a clear and bright appearance with proper color;

b. Within 31 days of addition of the new fuel oil to storage tanks verify that the properties of the new fuel oil, other than those addressed in

a. above, are within limits for ASTM 2D fuel oil; and

c. Total particulate concentration of the fuel oil is _<10 mg/I when tested every 92 days in accordance with the applicable ASTM standard.

d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

5.5.13 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

Point Beach 5.5-12 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Technical Specifications (TS) Bases Control Program (continued)

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 5.5.13b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.14 Safety Function Determination Procram (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

Point Beach 5.5-13 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power or loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.15 Containment Leakage Rate Testinq Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995.

b. The peak design containment internal accident pressure, Pa, is 60 psig.

c. The maximum allowable containment leakage rate, La at Pa, shall be 0.4% of containment air weight per day.

Point Beach 5.5-14 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate Testing Program (continued)

d. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is _*1.0 La.

2. During the first unit startup following testing in accordance with this program, the leakage rate acceptance are < 0.6 La for the combined Type B and Type C tests and 5 0.75 La for the Type A tests.

3. Air lock testing acceptance criteria are:

i. Overall air lock leakage rate is < 0.05 La when tested at _

Pa.

ii. For each door seal, leakage rate is equivalent to _*0.02 La at >_Pa when tested at a differential pressure of _ to 10 inches of Hg.

e. The provisions of SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Prbgram.

f. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Point Beach 5.5-15 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)

Leakage Program A program shall be established to verify the leakage from each RCS PIV is within the limits specified below, in accordance with the Event V Order, issued April 20, 1981.

a. Minimum differential test pressure shall not be less than 150 psid.

b. Leakage rate acceptance criteria are:

1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

5.5.17 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1990.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

Point Beach 5.5-16 Unit 1 - Amendment No. 226 Unit 2 - Amendment No. 206