ML20078P947
| ML20078P947 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/16/1994 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20078P951 | List: |
| References | |
| NUDOCS 9412210059 | |
| Download: ML20078P947 (4) | |
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SURVEllt ANCE REOUTREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current
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techniques prior to service to establish a baseline condition I
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.
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- 10) Ipbe Repair refers to a process that reestablishes tube i
serviceability. Acceptable tube repairs will be performed by j
the following processes:
l a)
Laser welded sleeving as described in a Westinghouse l
Technical Report currently approved by the NRC, subject to i
the limitations and restrictions as noted by the NRC staff, l
l or 1
b)
Kinetic welded sleeving as described in a Babcock & Wilcox Nuclear Technologies Technical Report currently aproved by the NRC, subject to the limitations and restrictions as noted by the NRC staff.
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Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure. A tube inspection per 4.4.5.4.a.8 is required prior t.o returning previously plugged tubes to service.
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- 11) Tube Support Plate Interim Plunaina Criteria limit for Unit 1 Cycle 5 is used for the disposition of a steam generator tube for continued service that is experiencing ODSCC confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging criteria limit, the tube's disposition for continued service will be based upon l
standard bobbin coil probe signal amplitude of flaw-like i
indications.
The plant specific guidelines used for all i
inspections shall be consistent with the eddy current guidelines l
in Appendix A of WCAP-13854 as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the voltage parameters as specified in Specification 4.4.5.2.
Pending incorporation of the voltage verification requirements in ASME standard verifications, an l
ASME standard calibrated against the laboratory standard will be l
utilized in Unit I steam generator inspections for consistent voltage normalization.
1.
A tube can remain in service with a flaw-like bobbin coil signal amplitude of less than or equal to 1.0 volt, regardless of the depth of the tube wall penetration, provided Item 3 below is satisfied.
2.
A tube can remain in service with a flaw-like bobbin coil j
signal amplitude greater than 1.0 volt but less than or equal to 2.7 volts provided an RPC inspection does not detect degradation and provided Item 3 below is satisfied.
9412210059 941216 PDR ADOCK 05000456 UNIT 1 -AMENDMENT N0. 55 BRAIDWOOD - UNITS 1 & 2 3/4 4-17 UNIT 2 -AMENDMENT NO. 54
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~ SURVEILLANCE REOUIREMENTS (Continued) s, 3.
The projected end of cycle distribution of crack indications is verified to result in total primary to secondary leakage less than 9.1 gpm (includes operational and accident leakage). The basis for determining expected 1eak rates from the projected crack distribution is provided in WCAP-14046, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria" dated May 1994.
4.
A tube with a flaw-like bobbin coil signal amplitude of '
greater than 2.7 volts shall be plugged or repaired.
Certain tubes identified in WCAP-14046, "Braidwood Unit-1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria," dated May 1994, shall be excluded from application of the tube support plate interim plugging criteria limit.
It has been determined that these tubes may collapse or deform following a postulated LOCA + SSE.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all i
tubes exceeding the plugging or repair limit) required by Table 4.4-2.
4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission 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 Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report 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 repaired.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
UNIT 1-AMENDMENT N0. 54 BRAIDWOOD - UNITS 1 & 2 3/4 4-17a UNIT 2-AMENDMENT NO. 54
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BASES I
l' 3/4.4.5 STEAM GENERATORS l'
The Surveillance Requirements for inspection of the steam generator tubes i
ensure that the structural integrity of this portion of the RCS will be main-tained. 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 surveil-lance 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 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.
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 i
chemistry is not maintained within these limits, localized corrosion may I
likely result in stress corrosion cracking.
The extent of cracking during i
plant operation would be limited by the l'aitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System l
(reactor-to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a reactor-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 reactor-to-secondary leakage of 150 gallons per day per i
l steam generator can readily be detected by radiation monitors of steam l
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 or repaired by sleeving. The technical bases for sleeving are described in the current Westinghouse or Babcock & Wilcox Nuclear Technologies Technical Reports.
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Wastage-type defects are unlikely with pro)er chemistry treatment of the I
secondary coolant. However, even if a defect s1ould develop in service, it I
will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit of 40% of the tube nominal wall thickness.
If a sleeved tube is found to contain a through wall penetration in the sleeve of equal to or greater than 40% of the nominal wall thickness, the tube must be plugged.
The 40% plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis and utilizes a 20% allowance for eddy current uncertainty and additional degradation growth.
Inservice inspection of sleeves is required to ensure RCS integrity. Sleeve inspection techniques are described in the current Westinghouse or Babcock & Wilcox Nuclear Technologies Technical Reports.
Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation that has penetrated 20% of the pressure retaining portions of the tube or sleeve wall thickness.
Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
UNIT 1 - AMENDMENT NO. 55 BRAIDWOOD - UNITS 1 & 2 8 3/4 4-3 UNIT 2 - AMENDMENT NO. 54
REACTOR COOLANT SYSJIM
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BASES l
3/4.4.5 STEAM GENERATORS (continued)
For Unit 1 Cycle 5. tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates will be dispositioned in accordance with Specification 4.4.5.4.a.ll, Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the.Ammission pur-suant to Specification 6.9.2 prior to resumption of plant operation.
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.
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I UNIT 1 - AMENDMENT NO. 50 BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 a UNIT 2 - AMENDMENT NO. 54
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