ML20043C033
| ML20043C033 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/25/1990 |
| From: | Larkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043C032 | List: |
| References | |
| NUDOCS 9006010223 | |
| Download: ML20043C033 (10) | |
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UNITED STATES 3
NUCLEAR REGULATORY COMMISSION f
WASHtNOTON O. C. 20666 PORTLAND GENERAL ELECTRIC COMPANY THE-CITY OF EUGENE. OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344
_ TROJAN NUCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.160 License No. NPF-1 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Portland General Electric Company, Ted Jan,ua(the licensee) dated November 30, 1989 as et al.
supplemen ry 25,1990 and A>ril 16 1990 com lies with the standards and requirements of tie Atomle Energy ket of 1954,asamended(theAct),andtheCommission'sregulations set forth in 10-CFp Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.
There is reasonable assurance (1) that the activities author-ized 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 Comission'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 &ccordance with 10 CFR 51 of the Comission's regulations and all applicable requirements have been satisfied.
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i 2.
Accordingly, the license is amended by changes to the Technical Specifi-i cations as indicated in the attachment to this license amendment, and paragrcph2.C.(2)ofFacilityOperatingLicenseNo.NPF-1ishereby amended to read as follows:
f (2) Technical Specifications The Technical Specifications contained in Appendices A and 0, as revised through Amendment No.160, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions, except where otherwise stated in specific ifcense conditions.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 4
John T. Larkins, Acting Director Project Directorate V l
Division of Reactvr Projects !!!,
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IV, Y and Special Projects L
Office of Nuclear Reactor' Regulation l
Attachment:
Changes to the Technical i
Specifications Date of Issuance: May 25, 1990
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l ATTACHMENT TO LICENSE, AMENDMENT NO.160 TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO. 50-344
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t Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages insert Pages 3/4 4-6 3/4 4-6 l
3/4 4-7 3/4 4-7 I
3/4 4-9 3/4 4-9 3/4 4-9a 3/4 4-9a 3/4 4-11 3/4 4 11 B3/4 4-2 B3/4 4-2 B3/4 4-2a B3/4 4-2a p
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STEAM GENERATORS LIMITING CONDITION FOR OPERATION m
3.4.5 Each steam generator shall be OPERABLE.
r APPLICABILITY: MODES 1, 2, 3 and 4.
ACTif3:
With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200'F.
SURVE!LLANCE REQUIREMENTS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam l
generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Sample Selection and inspection - The steam generator tube minimum sample size, inspection result classifi-cation, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes'in all
- l steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then e
at least 50% of the tubes inspected shall be from these critical creas, b.
The first sample of tubes selected for each inservice impection (subsequent to the preservice inspection) of each steam generator shall include:
1.
All tubes that previously had detectable tube wall penetrations
(>20%) that have not been plugged nor sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged.
1 2.
Tubes in those areas where experience has indicated potential problems.
j TROJAN-UNIT 1 3/4 4-6 Amendment No. 57, 160 j
REACTOR COOLANT SYSTEM SURVEILLANCE RE001REMENTS (Continued) 3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube.
If any selected i
tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an i
adjacent tube shall be selected and subjected to a tube inspection, c.
In addition to the 3% sample, all tubes with defects below the F* distance which have not been plugged shall be inspected in the tube sheet region, d.
The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include the tubes f rom those areas of the tube sheet array where tubes or tube sleeves with imperfections were previously found.
P.
The inspections include those portions of the tubes or tube sleeves where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Cateaorv Inspection Results C-1 Less than 5% of the total tubes inspected i
are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of-the total tubes inspected are degraded tubes.
C-3 More than 10% of the tots) tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations, i
TROJAN-UNIT 1 3/4 4-7 Amendment No. 57, 733, 160 l
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RACTOR COOLANT SYSTEM i
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceo,tance Criteria j
a.
As used in the Specification:
l 1.
Imperfection means an exception to the dimensions, finish or contour of a tube or tube sleeve f rom that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or tube sleeve wall thickness, if detectable, may be considered as imperfections.
2.
Dearadatio_q means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube or tube sleeve.
3.
pearaded Tube means a tube containing imperfections >20%
of the nominal tube or tube sleeve wall thickness caused by degradation.
4.
% Dearadation means the percentage of the tube or tube sleeve i
wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the repair limit.
A defective tube is a tube containing a defect that has not been repaired by sleeving or a sleeved tube that has a defect in the sleeve.
6.
Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the af fected area because it inay become unserviceable prior to the next inspection.
This definition does not apply to the area of the tube sheet region below the F* distance, provided the tube has no indications of cracking within the F* distance.
The repair limit imperfection depths are specified in percentage of-nominal wall thickness as follows:
a) Original tube wall 40%
b)
Babcock & Wilcox sleeve wall
- 40%
c)
Bechtel/KWU sleeve wall
- 35%
7.
ynserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its 1
structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater-line break as specified in 4.4.5.3.c above.
- The following sleeve designs have been found acceptable by the NRC staff:
1.
Babcock & Wilcox kinetic welded sleeves (BAW-2094P, Revision 1) 2.
Bechtel/KWU welded sleeves (BKA1-01-P, Revision 1; E0R-TRJ-01-P)
TROJAN-UNIT 1 3/4 4-9 Amendment No. 67, 733, 160
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SURVI!LLANCE REOUTREMENTS fContinued) l 8.
Iyfae !nsoection 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 9.
Preservice inspection means an inspection of the full length of each tube in each steam generator performed by eddy cunent techniques prior to service to establish a' baseline condition of the tubing.
This inspection shall be perfonned after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be i
used during subsequent inservice inspections.
1 10.
Tube Roll M ansion is that portion of a tube which has been increased in' diameter by a rolling process such that no
]
crevice exists between the outside diameter of the tube and the tube sheet.
11.
F* Distance is the minimum length of the roll expanded portion i
of the tube which cannot contain any. indications of cracking in order to ensure the tube does not pull out of the tube sheet.
The F* distance is 1.4 inches and is measured from the top of the roll expansion of the tube down toward the bottom of the tube sheet.
F* is not applicable if a sleeve has been installed in the roll expanded portion of the tube, b.
The steam generator shall be determined OPERABLE af ter completing the corresponding actions (plug or sleeve in the affected areas all tubes exceeding the repair limit) required by Table 4.4-2.
4.4.5.5 Reports a.
Following each inservice. inspection of steam generator tubes, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission. within 15 days.
b.
The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed.
This report shtO 1
include.
3 1
1.
Number and extent of tubes inspected.
j 2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
j 3.
Identification of tubes plugged or sleeved.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reviewed for reportability pursuant to l
Specification 6.6.1.
If the results are deemed reportable, such report must be submitted to the Commission prior to the resumption of plant operation.
TROJAN-UNIT 1 3/4 4-9a Amendment No. 57, 733, 160 i
--eg TABLE 4.4-2 STEAN GENERATOR TUBE INSPECTION
$iG 1ST SANPLE INSPECTIrN 2ND SANPLE INSPECTION 3RD SANPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A 5 Tubes per S.G.
C-2 Plug or sleeve C-1 None N/A N/A I
defective tubes and inspect additional Plog or sleeve C-1 None C-2 defective tubes and 3
25 tubes in this S.G.
C-2 Plug or steen inspect additional y
45 tubes in this S.G.
defective tubes Perfom action for w*
C-3 C-3 result of first l
a sample u
C-3 Perform action for N/A N/A C-3 result of first sample C-3 Inspect all tubes in All other this S.G., plug 5.G.s are None N/A N/A l
or sleeve C-1 N
Some S.G.s Perform action for N/A N/A and inspect 25 tubes C-2 but no C-2 ressit of second In each other S.G.
additional sample t
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g' Report to NRC e
pursuant to spec 1-Additional Inspect all tubes in E
fication 4.4.5.5.c.
S.G. is C-3 each S.G. and ping or 2
sleev: defective
,jg
,fg S
tubes. Report to NRC perssant-to spect-2
,o f1 cation 4.4.5.5.c.
w 5-3-% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected N
5 during an inspection.
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REACTOR (_00LANT SYSTE_M BASES iho power operated relief valves (PORVs) operate to relieve RCS pres-sure below the setting of the pressurizer code safety valves.
These relief valves have remotely operated block valves to provide a positive shutof f capability should isolation of a relief valve be necessary.
3/4.4.4 PREESURIZER The requirement that 150 kw of pressurizer heaters and their associ-ated controls be capable of being supplied electrical power f rom an i
emergency bus provides assurance that these heaters can be energized during a loss of of f site power condition to maintain natural circulation at HOT STANDBY. A minimum of 7 of the 23 kw heaters meets this requirement.
3/4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capa-bility to remove decay heat after a reactor shutdown.
The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RCS temperatures greater than 350*F if one steam gen-erator becomes inoperable due to single failure considerations.
Below 350'F, decay heat is removed by the RHR system.
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.
The program for Inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.63, Revision 1.
Inservice inspection of steem generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degra-dation 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 or tube sleeve 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 parameter limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant TROJAN-UN11 1 B 3/4 4-2 Amendment No. 58, 77d,160 t
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' REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAN GENERATORS CONTINUED system and the secondary coolant system (primary-to-secondary leakage =
500 gallons per day per steam generator).
Cracks having a primary-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 primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam 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 sleeved in the affected areas.
l Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.
However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving in the affected area will be required of all tubes with imperfections exceeding the repair limit which is defined in Specification 4.4.5.4.a.
Tubes with defects below the F*
distance do not have to be plugged or repaired as long as there are no indications of cracking in the F* distance.
The F* distance is'1.4 inches and includes a safety factor of 3 and a 0.5-inch eddy current measurement uncertainty.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage-type degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 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 l
tubes which are not degraded.
Descriptions of sleeve designs shall be sub-mitted to the NRC for review and approval prior to their use in the repair of degraded steam generator tubes.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 1
3/4.4.6.1 LEAKAGE DETECTION SYSTEMS l
The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.
These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.
TROJAN-UNIT 1 B 3/4 4-2a Amendment No. 57, 133, 160