ML20199H717
| ML20199H717 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/18/1997 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199H722 | List: |
| References | |
| GL-95-05, GL-95-5, NUDOCS 9711260205 | |
| Download: ML20199H717 (32) | |
Text
. ___ ______-_
- "Cug lt UNITE 3 STATES g:
,j NUCLEAR REGULATORY COMMISSION
'a WASHINGTON, D.C. DetNM001
,s,*..+/
NORTHERN STATES POWER COMPANY DOCKET NQM2 PRAIRIE ISLAND NULLEAR GENERATING PL_ ANT. UNIT NO 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.133 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northem States Power Company (the licensee) dated May 15,1997, as supplemented August 29, October 20, October 24, and October 28,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, ar'd (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requ'rements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as Indicated in the attachment to this license amendment, ar'd paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No. DPR-42 are hereby amended to read as follows:
9711260205 971118 PDR ADOCK 05000282 P
2
.(2)
Technical Speedications The Technical Specifications coMained in Appendix A, as revised through Amendment No.133, are herooy incorporated in the license. - The licensee shall operate the facility in accordance with the Technical Specifications.
(5)
Addnional Conddions The Additional Conditions contained in Appendix B, as revised through Amendment No.133, are hereby incorporated into this license. The lic4,nsee shall operate the facility in accordance with the Additional Conditions.
3.
This license amendment is effective as of the date of issuance, with full implementation of the Technical Specifications within 30 days.- License Condition 5 of Appendix B shall be implemented immediately upon issuance of this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION h, Y, Beth A. Wetzel, Senior Project Manager Project Directorate ill 1 Division of Reactor Projects -Ill/IV Office of Nuclear heactor Regulation
Attachment:
- 1. Changes to de Technical Specifications
- 2. Appendix B Additional Conditions Date of lasuance: November 18, 1997 i
5 V
AT ?ACHMENT TO LICENSE AMENDMENT N3.133 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50 282 Revise Appendix A Technical Specifications by removing the pages identified below and
[
inserting the attached pages. The revised pages are identified by amendment number and contain verticallines indicating the area of change.
REMOVE litSERT TS vi TS vi TS.3.1 9 TS.3.1 9 TS.4.12 3 TS.412 3 TS.4.12-4 TS.4.12-4 TS.4.12 5 TS.4.12 5 TS.4.12 6 TS.4.12 6 TS.4.12 7 TS.4.12 7 B.3.1-7 B.3.1-7 B.4.12 1 8.4.12 1 B.4.12 2 B.4.12 2 B.4.12 3 B.4.12 3 B.4.12 4 B.4.12 4
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TS vi I61LE OF CONTENTS (Continued)
TS SECTION IIILE PAGE 4.12 Steam Cenerator 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 Inspectien C. Inspection Frequencies TS.4.12 3
- b. 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 1 4.16 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 J
Prairie Island Unit l' Amendment No. 122, 129F. 133 Prairie Ialand Unit 2 Amendment No. LFT. 12T. 125
i TS.3.lo9 l
r 3.1.C.2
- e. If the total reactor coolant system to secondary coolant system
- leakage through any one steam generator of a unit exceeds 150-gallons per day (GFD), within one hour initiate action to place the 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 be in CUI.D SHUTDOW within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and perform an inservice steam generator tube inspection in i
accordance with Technical Specification 4.12.
3.
Pressure Isolation Valve tankane 14akage through the pressure isniation valves shall not exceed tha
- maximum allowable leakage specified in Specification 4.3 when--
+
. reactor coolant system average temperature exceeds 200*F.
If the maximum allowable leakage is exceeded, within one hour initiate the action necessary to place the 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 />.
i
. Prairie Island Unit 1 Amendment No. P(, 133 Prairie Island Unit 2 Amendment No JK 125
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i TS.4.12 3 i
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- 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.
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- 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.
Ina21stion Freauencies.1he above required in. service inspections of steam generator tubes shall be perforced 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 no: 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 stets 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 a 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.co tube sheet welds) in excess of the limits of Specification 3.1.C.6.
(b)
A seismic occurrence greater 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.
Prairie Island Unit 1 Amendment No. )di 133 Prairie'Islan6 Unit 2 Amendment No J27, 125
TS.4.12 4 i
D.
Accentance criteria 1.
As used in this Specification:
(a)
Innerfection means an exception es 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 thickaess, if detectable, may be considered as imperfections.
(b)
Derradation means a service. induced cracking, wastage, wear et general corrosion occurring on either inside or outside of a tube.
(c)
Derraded Tube means a tube containing imperfections 120% of the nominal wall thickness caused by degradation.
(d) t 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)
Renair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by slesving 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 poecion 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 dcas 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 applicable 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 Insceetion 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)
Sleevinz 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 tubesheet sleeve with a hard rolled joint.
The new cleeve becomes the pressure boundary spanning the original degraded tube region.
Prairie Island Unit 1 Amendment No. JJer. k37,133
- Pralrle Island Unit 2 Amendment No..111,,124, 125 i
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F* Distance 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 current uncertainty).
(b)
F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, 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 limit and all tubos contain'.g through wall cracks or classify as F* tubes) required by Ta' is TS.4.12 1 and TS.4.12 2.
3.
Tube repair, after October 1, 199 asing Co.nbustion Engineering welded sleeves shall be in accordan, with the methods described in the following:
CEN-629 P, Revision 2, " Repair of Westinghouse Series 44 and $1 Steam Generator Tubes Using Leak Tight Sleeves";
CEN 629 P, Addendun 1 Revision 1, " Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves" 4
Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predouinantly 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 st ress 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 servt.e if a rotating pancake coil (or comparable examination technique) inspection does not detect degradation.
Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired.
Prairie Island Unit 1 Amendment No, M6' DC 133 Prair
- Island Unit 2 Amendment No. JK, J2#; 125
TS.4.12 6 i
- d. If an unscheduled mid. cycle inspection is performed, the following mid. cycle repair limits apply instead of the limits in j
Specifications 4.12.D.4.a. b and c.
The mid. cycle repair limits are determined from the following equations Y"
V'a=
1.0*NDE+G t
-i Vau=V m -(Vac3.0)
'l where:
Vunt - upper voltage repair limit V at - lower voltage repair limit
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- Vm - mid. cycle upper voltage repair limit based on time into cycle V m mid. cycle lower voltage repair limit based on V m and r
time into cycle ot - length of time since last scheduled inspection during which Vumt and V at were implemented t
CL = cycle length (time between two scheduled steam generator i
inspections)
Vit - structural limit voltage Cr - 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 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 1.etter 95 05 as supplemented.
t 1
Prairie Island Unit 1 Amendment No. 133
- Prairie lsland Unit 2 Amendment No. 125
T5.4.12 7 E.
]Ltoorts
- 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 Comnission 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 generator 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 steam generator tube inspections which fall into Category C 3 require notification to the Commission prior to resumption of plant operation, 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 taken to prevent recurrence.
- 4. The results of inspections performed under Specification 4.12.B for all tubes that have defects below the F* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:
a.
Identification 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 steam generators to service should any of the following conditions arise:
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 tne tube support plate intersections, c.
If 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 attributable to primary water stress corrosion cracking.
e.
If the calculated conditiona. burst probability based on the projected end of cycle (or it not practical, using the actual measured end-of cycle) voltage distribution exceeds 1 x 10-8, notify the NRC and provide an assessment of the safety significance of the occurrence.
Prairie Island Unit 1 Amendment No. Mff, J37,133 Prairie Island Unit 2 Amendment No. 1 W,,R r, 125
B.3.1 7 3.1 MACTOR COOLANT SYSTIM Bases continued C.
Reactor Coolant System Leakage Leakage from the reactor coolant system is collected in the containment or by other systems. These systems are the main steam system, conden.
sate and feedwater system and the chemical and voluma control system.
Detection of leaks from the reactor coolant system is by one or more of the following (Reference 1):
1.
An increased amount of makeup water required tw maintain normal level in the pressurizer.
2.
A high temperature alarm in the leakoff piping provided to collect reactor heac' flange leakage.
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 monitors 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.
Steam 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 assuranca that should a significant leak be experienced in service, it will 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 piping.
RefJtrences 1.
USAR, Section 6.5 2.
USAR, Section 7.5.1 3.
Testimony by J Knight in the Prairie Island public hearing on January 28, 1975, pp 13 17.
Prairie Island Unit 1 Amendment No. STl J06,-133
' Prairie Island Unit 2 Amendment No. A4, 97, 125
B.4.12 1 4.12 STEAM Cf.NERATOR TURE SURVEILIANCE Aaaaa The Surveillance Requirements for insp d on of the steam generator tubes i
ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator cubes is based on a modification of Regulatory Guide 1.83. Revision 1.
l In. service inspection of steam generator tubing is essential in order to maintain surveillance of the condittons of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, uanufacturing 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.
At the request of the NRC, a requirement for in service inspection of at least 20% of the total number of sleeves 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 sleeve. Type 6 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 or not the welded joints have 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 primary 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 failure due to loads imposed by design basis accidents (Reference 1),
Operational experience has demonstrated that primary te. 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 stea:n generator will require plant shutdown and aa unscheduled eddy current in2pection, during which the leaking tubes will be located and plugged or sleeved.
Wastage type defects are unlikely with proper chemistry treatment of r
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 accaptable wall thickness prier to the next inservice inspection which, by the definition
.of Specification 4.12.D.1.(f). is 50%'of the tube or sleave nominal wall thickness. Wastage type defects having a wall thickness greater than Prairie Island Unit 1 Amend nent No.,9f, JM,133 Prairie' Island Unit 2 Amendment No. ) (, J M, 125
B.4.12 2 4.12 STEAM CENERATOR TUBE SURVEIL 1HICE Bases continued 0.025 inches will have adequat-margins of safety against failure due to loads imposed by normal plant operation and design basis accidents (Referar.co 1).
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type defects that have penetrated 20n of the original 0.050 inch wall thickness (Reference 2).
Plugging or sleeving is no t required for tubes meeting the F* criteria.
The F* distance will be co. trolled by a combination of eddy current inspection er,J/or procese 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 effective 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 inch above the original roll expansion.
In the case of "se new roll, eddy current examination will confirm there are no indications in the r.ew roll region and that there is a new roll region with well defined upper and lower expansion transitions.
When eddy current examination, alone, must determine the F* distance, such as in the existing hard roll region, or when multiple lengths 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 cur? int (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 summation 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 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.
Degraded steam generator tubes may be repaired by the installa Jon of sleeves which span the section of degraded steam generator tubing. A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded.
The following sleeve designs have been found acceptable by the NRC Staff;
- a. Westinghouse Mechanical Sleeves (WCAP 10757)
- c. Combustion Engineering Leak Tight Sleeves (CEN 294 P, for sleevis installed prior to October 1, 1997)
- d. Combustion Engineering Leak Tight Sleeves (CEN 629-P, for sleeves installed after October 1, 1997)
Prairie Island Unit 1 Amendment No. JJ6F,137',133 Prairie Island Unit 2 Amendment No. J14', 124', 125
3.4.1203 4.12 JTEAM CENERATOR WBE SURVEILIANCE Bases 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 Support Plate Repair Limit 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 outside 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 applicable to ODSCU that occurs at other locations within the steam generator. Additionally, ?.he repair criteria apply only to indications where the degradation mechanism is dominantly axial CDSCC with no significant cracks extending outside the thicknest of the support plate.
Refer to Generic Letter 95 05 for Odditional 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 voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltaga 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 material properties at 650'F (i.e., the 95 percent LTL curve).
The voltage 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
- Vms where Va, represents the allowance for flaw growth between inspections and Vma represents the allowance for potential sources of erro in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit 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.
Prairie Island Unit 1 AmendmentNo.;9([J41,133 Prairie Island Unit 2 Amendment No.
4', JJf, 125 l
B.4.12 4 4.12 STEAM GENERATOR TUBE SURVEILLANCE 113.g1 continued Specification 4.12.E.5 implements several reporting requirements recommended by Generic Letter 95 05 for situations which the NRC vants 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 Generic Lettor 95 05 for more information) when it is not practical to complete these calculations using the projected EOC volta 6e distributions prior to returning the steam generators to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the -trposes of addressing Generic Letter 95 05 Section 6.a.1 and 6.a.3 repot'.ig criteria, then the results of the projected EOC voltage distribution should be provided per Generic Letter 95 05 Section 6.b (c) criteria.
Refere. a 1.
Testimony of J Knight in the Prairie Island Public Hearing on 1/28/75 2.
Testinony of L Frank in the Prairie Island Public Hearing on 1/18 Prairie Is/75 land Unit 1 Amendment No. A 6 J16 133 Prairie Island Unit 2 Amendment No. f 4.llI, 125
APPENDIX B ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. DPR-42 Northern States Power Company shall comply with the following conditions on the schedules noted below:
~ Amendment implementation Number Adddional Conddion Date 128 1, NSP will provide a licensed operator in the control room Prior to Unit 2 entering on an interim basis for the dedicated purpne of identifying Mode 2.
an earthquake which results in a decreasing ogfeguards cooling water bay level. This operator will be,.n addition to the normal NSP administrative control room steffing
~
requirements and will be provided until Lkense Condition 2 is satisfied.
128
- 2. NSP will submit dynamic finite element analyses of the July 1,1997, and intake canal banks by July 1,1997 for NRC review. By December 31,1998, December 31,1998, NSP vall complete, as required, as stated in Condition additional analyses or physical modifications which provide 2.
the basis for extending the time for operator post-seismic cooling water load management and eliminating the dedicated operator specified in License Condition 1.
128
- 3. Based on the results of License Condition 2, NSP will At the next USAR revise the Updated Safety Analysis Report to incorporate update following the changes into the plant design bases. These changes completion of will be included in the next scheduled revision of the Condition 2, but no Updated Safety Analysis Report following compleUon of later than June 1,
- License Condition 2 activitiesi 1999.
130'
- 4. Prairie Island will assure that heavy loads do not present This is effective a potential for damaging irradiated fuel through use of 1) a immediately upon single-failure-proof crane with rigging and procedures which issuance of the
- Implement Prairie Island commhments to NUREG-0612; or amendment.
- 2) spent fuel pool covers with their implementing plant procedures for installation aruf use.
133
- 5. NSP will assure that during the implementation of steam This is effective generator _ repairs utilizing the voltage-based repair criteria, immediately upon the total calcu!sted primary to secondary side leakage from issuance of the the faulted steam generator, under main steamline break amendment.
7 conditions (outside containment and upstream of the main -
x steam isoU.icn valves), will not exceed 1.42 gallons per minute (based on a reactor cooiant system temperature of j~
_578 'F).-
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Amendment No.133
2034 y
i UNITED STATES g
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NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 200eH001
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.125 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated May 15,1997, as supplemented August 29, October 20, October 24, and October 28,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CcR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C,
There is reasonable assurance (l) thEt the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D,
The issuance of this amendment will not be inimical to the coinmon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's reguiations and all applicable sequirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No. DPR-60 are hereby amended to read as follows:
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2-s
.(2)-
Technical Specifications
- The Technical Specifications contained in Appendix A, as revised through
' Amendment No.124 pre hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(5)
Addhnnal Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No 125, are hereby incorporated into this license. The licensee-shall operate the facility in acwidance with the Additional Conditions.
3.
This license amendment is effective as of the date of issuance, with full implementation.
of the Technical Specifications within 30 days'. License Condition 5 of Appendix B shall be implemented immediately upon itauance of this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION 0.'h tb Beth A. Wetzel, Seni$r Project Manager Project Directorate lil ;
Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regu!ation Attachment -
- 1. Changes to the Technical Specifications
- 2. Appendix B - Additional Conditions
- Date of lasuance: November 18, 1997 4
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P ATTACHMENT TO LICENSE AMENDMENT NO.125 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50 306 Revise Appendix A Technical Specifications by removing the pages identified below and -
inserting the attached pages. The revised pages are identified by amendment number and contain vert:callines indicating the area of change.
REMOVE INSERT
. TS-vi TS-vi
- TS.3.1 9 TS.3.1 9 TS.4.12 3 TS.412 3 TS.4.12-4 TS.4.12-4 TS.4.12-5 T8.4.12-5 TS.4.12-6 TS.4.12-6 TS.4.12-7 TS.4.12 7 B.3.1-7 B.3.1-7 B.4.12-1 B.4.12-1 B.4.1 *e-2 B.4.12-2 B.4.12-3 B.4.12-3 B.4.12-4 B.4.12-4
TS vi TABLE OF CONTENTS (Continued)-
TS SECTION IIILE PACE 4.12 Steam Generator Tube Serveillance TS.4.12 1 A. Steam Generator Sample Selection and TS.4.12 1 Inspaction 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 1 4.16 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 Prairie Island Unit 1 Amendment No. 122, 127, 133-Prairie Island Unit 2 Amendment No. LF5",12T,125 l
TS.3.1-9 3.1.C.2
- e. If the total reactor coolant system to secondary coolant system leakage through any one steam generator of a unit exceeds 150 Sailons per day (GPD), within one hour initiate action to place the unit in HOT SHUTDOWN and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD CHUTDOWN within t'.ie following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and perform 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 manimum allowable leakage specified in Specification 4.3 when reactor coolant system average temperature exceeds 200*F.
If the maximra allowable. leakage is exceeded, within one hour initiate-the action necessary to place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN 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 SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Prairie Island Unit 1 Amendment No, jd',133 Prairie Island Unit 2 Amendment No. JWP,125
TS.4.12 3 I
S. Indicationt 3ft $n service as a result of app 14 cation of tube support plate volt s-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 sapport plate with bown 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 e 20 percent random sampling of tubes inspected over their full length.
C.
Insoection 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 a maximum of once per 40 months,
- 3. Additional, unscheduled inservice inspections shall be performed on each steam v nerator 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 ori inating C
from tube to-tube sheet welds) in excess of the limits of Specification 3.1.C 6.
(F )
A seismic occurrence greater than the Operating Basis Earthqus'.:e.
(c)
A loss-of coolant accident requiring actuation of the engineered safeguards.
(d). A main steam line or feedwater line blaak, i
Prairie Island Unit 1 Amendment No. X 133 Prairie Island Unit 2 Amendment No. )5',125 l
TS.4.1204
)
D.-
-. Accanemnee Criteria j
1.
As used in this-Specification:--
(a).
Innerfection means an exception to the dimensions, finish or l
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.
a considered as imperfections s (b)
Dearadation means a service. induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a-tube.
'(c)
Degraded Tube means a tube containing imperfections a:20% of the
_ nominal wall thickness caused by degradation.
(d) t Derradation means the percentage of the tube wall thickness affected or removed by degradation.
-(e)
Defect means an imperfection of such sevexity that it exceeds the repair limit. A tube containing a defect is defective.
(f)
'Renair 1imit means the imperfection depth at or beyond which the tube shall be_ removed from service by plugging or repaired i
by sleeving because it may become enserviceable prior to the next inspection and is equal to 50% of the nominal tube wall J
thickness.
If significant general tube thinning occurs, this criteria vill 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 degra6ed (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 applicable to these intersections.
(g)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its struerural r
integrity in the event of an Operating Basis Earthquakt, a
' loss-of coolant accident, or a steam line or feedwater line break.
(h)
Tube Insnection 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)
Sleevinn is the repair.of degraded tube regions using,a new A11oy 690 tubing sleeve-inserted inside the parent tube and sealed at each end by welding or by replacing the lower weld in 4
a full depth tubesheet sleeve with a hard rolled joint. The new-sleeve becomes the pressure boundary spanning the original degraded tube region.
' Prairie _ Island Unit 1-Amendment No. p8", El, ' 133 Prairie Island. Unit 2' Amendment No. IM, E, 125
TS.4.12 5 l
(j)
F* Distance _is the distance from the botton'of the hardroll-
-t
- E
-transition-toward the bottom of the tubesheet that has been i
conservatively determined to be 1.07 inches-(not including eddy
' current uncertainty).
1(k)_-
Z*2dta is a tube with. degradation, _belew the F* distance, 4
equal _to or greater than 404, and not degraded (i.e., no indications of cracking) within the F* distance.
2.-
The steam generator: shall be determined OPERABl.E after completing the
= corresponding actions (plug or repair by sleeving all. cubes exceeding the repair limit and all tubes containing through wall cracks or-classify as F* tubes) required by Tables TS.4.12 1 and TS.4.12 2.
3.
Tube repair, after October 1, 1997, using combustion Engineering welded sleeves shall be in accordance with the methods described in p
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 Sleeves" 4.-
Tube Sunnort Plate Renair 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 i;
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 generstor tubes, with indications - of potential degradation attributed to outsife diameterLatress 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-
' diameter stress corrosion cracking degradation with a hobbin
-voltage greater-than the upper voltage repair limit will be
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plugged or repaired, i-Prairie Island' Unit 1 Amendment No. K 13C 133
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~Prair$e Island Unit 2 Amendment No. JE, J3dr," 125
- 2....
TS.4.12 6
- d. If an unscheduled mid-cycle inspec. tion is performed, 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 are determined from the following equations:
Ya y
AC 1.0+NDE+G vm=va -(Vm-2.0)
Acl where:
Vn:. - upper voltage repair limit V m - lower voltage repair limit V m - mid cycle upper voltage repair limit based on time into cycle V u - mid cycle lower voltage repair limit based on V m and time into cycle at - length of time since last scheduled inspection during which Vaz, and Vm were implemented CL - cycle length (time between two scheduled steam generator inspections)
Vst. - 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 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.
Prairie Island Unit 1 Amendment No. 133 Prairie Island Urit 2 Amendment No. 125
TS.4.12 7 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 Commission within 15 days.
- 2. The results of steam generator tube inservice inspections shall be included with the summary reports of ASKE Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tuba inservice inspections not associated with a refueling outage shall be submitted wi-hin 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 steam generator tube inspections which fall into Category C 3 require notification to the Commission prior to resumption of plant operation, 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 taken to prevent recurrence.
- 4. The results of inspections performed vader Specification 4.12.B for all tubes that have defects below the F* distance, and were not plugged, shall be reported to the Commission within 15 days follcwing the inspection. The report shall include:
a.
Identification 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 steam generators to service should any of the following conditions arise:
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.
If indications are identified that extend bsyond the confines of the tube support plate, d.
If indications are identified at the tube support plate elevations that are attributable 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 eno of-cycle) voltage distribution exceeds 1 x 10 2, notify the NRC and provide an assessment of the safety significance of the occurrence.
Prairie Island Unit 1 Amendment No. Its, JLET 133 Prairie Island Unit 2 Amendment No. 311',30., 125
B.3.1 7 3.1 SEACTOR_LOQIM SYSTDi RA111 continued C.
Reactor Coolant System Leakage Leakage from the reactor coolant system is collected in the containment or by other systems. These systems are the main steam systs, conden-sate and feedwater system and the chemical and volume coe'-ol system.
Detection of leaks from the reactor coolant system is by one or more of the following (Reference 1):
1.
An increased amount of makeup water required to maintain normal level in the pressurizer.
2.
A high temperature alarm in the leakoff pipint. provided to collect reactor head flange leakage.
3.
Containment sump water level indication.
4 Containment pressure, temperature, and humidity indication.
If there is significant radioactive contaaination of the reactor coolant, the radiation monitoring system provides a sensitive indica-tion of primary system leakage. Radiation uonitors 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 l
crack less than 0.6 inches long bas 6d on test data.
Steam 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 leaka5e limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected, and the plant shut dovn in a timely manner.
Specifica: ion 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 piping.
References 1.
USAR, Section 6.5 2,
USAR, Section 7.5.1 3.
Testimony by J Knight in the Prairie Island public hearing on January 28, 1975, pp 13 17.
' Prairie Island Unit 1 Amendment No. 917 JAMf,133 FA 81rle Island Unit 2 Amendment No. ##, 91, 125
-B.4.12 1T l
4.12 STEAM GENERATOR TURE SURVEILIANCE T
L SAAAA The Surveillance Requirements for inspec' tion _of_the' steam generat_or tubes ensure that the structural integrity of this portion of the RCS will be maintained.- The program for inservice-inspection-of ateam generator > tubes is_ based on a modification of Regulatory Guide 1.83, Revision 1.
In service inspection of staan generator tubing is essential in order to 4
maintain-surveillance of the conditions of the tubes in the event that.
ithere~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 degrAcacion so.
that corrective measures can be taken.
At the request of the NRC, a requirement for in service it 'ection of at least 20% of the total number of sleeves in service in botn steam generators was added to TS 4.12.8.
In addition,_ Table TS 4.12 2 was added to p: ovide 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 Ge S iines. The sample selection is applied to each type of. sleeve._ Types et 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 or not the welded joints have
. received pust weld heat treatment.
The plant is expected o 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
- orrosion 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 primary coolant system and the secondary coolant system (primary-to secondary leakage - 150 1
gallons per day from one staan, generator). Historically, cracks not
-addressed by voltage based alternate repair criteria and having a primary to secondary leakage less than 1.0 gpa (1440 gallons per day) during operation would have~an adequate margin of. safety against failure s
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.
4-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 oddy current inspection, during which the leaking tubes will be located and plu5ged or sleeved, i
I
-Wastage 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 imperfeetions that c
~ could develop defects having-less-than the minimum acceptable wall thickne'ss 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
-Prairie Island' Unit l'~
Amendment No. ff, JM, _133 Prairie Island' Unit 2 Amendment No. JW, JM,125 -
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B.4.12 2
- 4.12 STEAM GENERATOR TURE SUL7EILIANCE Bassa-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 type defects that have penetrated'20% of the-original 0.050 inch _ wall thickness (Reference 2).
Flugging or sleeving is not required for' tubes meeting the F* criteria.
The-F* distance will be controlled by a-combination of eddy current
-inspection and/or process control. Fvr 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 effective 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 inch above the original roll expansion.
In the case of the new-roll -eddy current examination-will confica there are no indic tions in the new roll region and that there is a new roll region with well defined upper and
.Iower expension transitions.
When oddy current examination, alone, must determine the F* distance, such as in the existing hard roll region, or when multiple lengths of additional hard rolls have been added, the eddy current uncertainty is qualified by testing against known standards. That value is expected to
'>e 0.18-inches. Therefore, the F* distance seasured by oddy current (sua of 1.07 and 0.18) will be conservatively-set _at 1.3 inches.
i When more than one Alternate Repair Criteria are used, the summation 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 Alterna;;e 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 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.
' Degraded steam generator tubes may be repaired by the installation of sleeves which span-the section_ of degraded steam generator tubing. A
- steam generator tube with a sleeve installed meets the structural requirements.of tubes.which ara not degraded.
The:following sleeve designs have been found acceptable by the NRC Staff:
a.-Weatinghouse Mechanical Sieeves (WCAP-10757)
- b. Westinghouse Brazed-Sleeves (WCAP 10820) ic. Cosbustion Engineering Leak Tight Sleeves _(CEN-294 P, for sleeves-installed prior to October 1, 1997) d.. Combustion Engineering Leak Tight-Sleevas (CEN 629-P, for sleeves installed after October.1, 1997)
Prairie'IslandIUnit 1 Amendment No. JW, 16, > 133 Prairie' Island Unit 2 Amendment No. J M', P4', 125
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B.4.12 3 4.12-STEAM CENERATOR TUBE SURVEILIANCE
&aitt-continusd Descriptiens_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 Suncort Plate Renair 1Amit 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 outside 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 applicable to ODSCC that occurs at other locations within the steam generator. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominently ani:1 ODSCC with no sigtificant cracks extending outside the thickness of the support plate. Refer to Generic Letter 95-05 for add {tignaldescriptionofthedegradationmorphology.
Implementation of Specification 4.12.D.4 requires a derivation of the voltage structural.1.imit from the burst versus voltage empirical
. correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95 percent prediction interval curve reduced tn account for the lower 95/95 percent tolerance bound for tubing material properties at 650*F (i.e., the 95 percent LTL carve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for ND3 uncertainty.
The upper voltage repair limit; Vma, is determined from the structural vcitage limit by applying the followinC equation:
Vme - Vst - Vor
- Ver where Va, represents the allowance for flaw growth between inspections and Vmm represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit 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 addy current
~a is acquired for indications at the tube support plates.
Prairie Island Unit 1 Amendment No. ;pr[ JJ1[ 133 Prairie Island Unit 2 Amendment No. J4, JJ4'.125
B.4.12 4 4.12 STEAM CENERATOR TUBE SURVEILIANCE 1333A continued Specification 4.12.E.5 implements several reporting requirements recommended by Generic latter 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 fot.nd voltage distribution rather than the projected end of cycla voltage distribution (refer to Generic 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 generators to service. Note that if leakage and conditional burst probability were calculated using the measured EOC voltsge distribution for the purposes of addressing Generic Letter 95 05 Section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should i.,e provided per Generic Letter 95 05 Section 6.b (c) criteria.
References 1.
Testimony of J Knight in the Prairie Island Public Mearing on 1/28/75 2.
Testimony of L Frank in the Prairie Island Public Hearing on Prairie Island Unit 1 Amendment No. M % 133 Prairie Island Unit 2 Amendment No, SC.311,125 l
APPENDIX B f
ADDITIONAL CONDITIONS FACILITY OPE 8ATING LICENSE NO. DPR-60 4
h l
Northem States Power Company shall comply with the following conditions on the sc edu es noted below:
Amendment Implementation Number Additional Condition Date
.120
- 1. NSP will provide a licensed operator in the control room Prior to Unit 2 entering on an interim basis for the dedicated parpose of identifying Mode 2.
an earthquake which results in a decreasing safeguards cooling water bay level. _ This operator will be in addition to the normal NSP administrative control room staffing requirements and will be provided until License Condition 2 is satisfied.
120 -
- 2. NSP will submit dynamic finite element analyses of the July 1,1997,.and intake canal banks by July 1,1997 for NRC review. By December 31,1998, December 31,1998, NSP will complete, as required, as stated in Condition additional analyses or physical modifications which provide.
2.
the basis for extending the time for operator post-seismic cooling water load management and eliminating the dedicated operator specified in License Condition 1.
120
- 3. Based on the results of License Condition 2, NSP will At the next USAR revise the Updated Safety Analywis Report to incorporate update following the changes into the plant dasign bases. - These changes completion of will be included in the next scheduled revision of the Condition 2, but no Updated Safety Analysis Report following completion of later than June 1, License Condition 2 activities.
1999.
122
- 4. Prairie Island will assure that heavy loads do not present This is effective a potential for damaging irradiated fuel through use of 1) a immediately upon single-failure-proof crane with riggiag and procedures which issuance of the
. Implement Prairie Island commitments to NUREG-0612; or amendment.
- 2) spent fuel pool covers with their implementing olant procedures for installation atKi use.
~
125 5.' NSP will assure that during the implementation of steam - This is effective.
generator repairs utilizing the voltage-based repair criteria, immediately upon
. the total calculated primary to secondary side leakage from issuance of the the faulted steam generator, under main steamline break amendment.
conditions (outside containment and upstream of the main steam isolation valves), will not exceed 1.42 gallons per minute (based on a reactor coolant system temperature of f 78 'F).
Amendment No.125 z
2.
_