ML19347E954
| ML19347E954 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/08/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347E953 | List: |
| References | |
| NUDOCS 8105140407 | |
| Download: ML19347E954 (14) | |
Text
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION
, N.s) ;y p
WASHINGTON,0. C. 20555 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET H0. SC-344 M AN NUCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 License No. NPF-1 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Portland General Electric Company, the City of Eugene, Oregon, and Pacific Power and Light Company (the licensee) dated April 22, 1977, as. amended March 20,1979 and supplemented ' December 28, 1979 and February 5,1981, 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 confomity 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, and (ii) that such activities will be conducted in compliance with the Cummission'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 Tart 51 of the Commission'.s regulations and all applicable requi*ements have been satisfied.
97
>810514 0 i
l
2-2.
Accordingly, the ifcense is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-1 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 61, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, except as noted in paragraphs 2.C.(10) through 2.C.(12) below.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY. COMMISSION Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 8,1981
'GE e
ATTACHMENT TO LICENS$ AMENDMENT AMENDMENT NO. 61 TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO. 50-344 Revise Appendix A as follows:
Remove Paces Insert Pages 3/4 0-2 3/4 0-2 3/4 0-3 3/4 4-6 3/4 4-6 3/4 4-7 3/4 4-7 3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-9a 3/4 4-9a 3/4 4-29 3/4 4-29 3/4 4-30 through 3/4 4-49 B 3/4 0-5 B 3/4 4-2a B 3/4 4-2a B 3/4 4-9 B 3/4 4-9 B 3/4 4-10
%g e
1 u-.
a 3/4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) b.
A total maximum combined interval time for any 3 consecutive surveillance internals not to exceed 3.25 times the specified surveillance interval.
4.0.3 Performance of a Surveillance Requirement within the specified time interval shall ccr.stitute ccepliance with OPERABILITY requirements for a Limiting Condition for Operation ano associated ACTION statements unless otherwise required by the specification.
Surveillance Require-ments do not have to be performed on incperable equipment.
4.0.4 Entry into an CPERATIONAL MODE or other specified apolicability condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. *
+
The provisions of Specification 4.0.4 are not applicable to the perform-ance of surveillance activities associated with fire protection technical specifications 4.3.3.7.1, 4.3.3.7.2, 4.7.8.1.1, 4.7.8.1.2, 4.7.8.1.3, 4.7.8.3 and 4.7.9 until the completion of the initial surveillance interval associated with each specification.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:
a.
Inservice inspection of ASME Code Class 1, 2 and 3 compo-nents and inservice testing of ASME Code Class 1, 2 anc 3 pumps and valves shall be performed in accordance with Section XI of the ASME Soiler and Pressure Vessel Code and applicable Adcenda as required by 10 CFR 50, Sec-tion 50.55a(g), except wnere specific written relief has been granted by the Commissicn pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
b.
Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice testing activities required by the ASME Boiler and Pressure Yessel Code and applicable Addenda shall ce applicable as follows in these Technical Specifications:
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i TROJAN-UNIT 1 3/4 0-2 Amendment No. 61
3/4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)
ASME Boiler and Pressure Vessel Code and applicable Required frequencies Addenda terminology for for performing inservice inservice testing activities testing activities Weekly At least once per 7 oays Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days..
c.
The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
d.
Performance of the above inservice inspection and ' testing activities shall be in addition to other specified Surveil-lance Requirements.
e.
Nothing in the ASME Boiler and Pressure Yessel Code shall be construed to supersede the requirements of any* Technical Specification.
TRGJAN-UNIT 1 3/40-3 Amendment tio. 61
j REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable generator (s) to CPERABLE status prior to increasing T,yg above 200*F.
SURVEILLANCE REQUIREMENTS 4.4.5.1. Steam Generator Samole Selection and Insoection - Each stemn generator snall De cetermined CPERABLE curing shutcown Dy selecting and' inspecting at least the minimum numcer of steam generators specified in Tabl e 4.4-1.
4.4.5.2 Steam Generator Tube Samofe Selection and Inspection - The steam generator tuae minimum sample size, inspection result classiff-cation, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be perfomed 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 l
insection shall include at least 3% of the total numoer of tubes in all 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 at least 50*. of the tubes inspected shall be from these critical areas.
b.
The first sample of tubes selected for each inservice inspection (subsequent to tne preservice inspection) of each steam generator shall include:
I i TRCJAN-UNIT 1 3/4 4-6 Amendment tio. 61
[
, SURVEILLANCE REQUIREMENTS (Continued)-
1.
All nonplugged tubes that previously had detectable wall penetrations (>20%).
2.
Tubes in those areas where experier:e has indicated potential problems.
3.
A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be perfomed cn each selected tube.
If any selected tube does not pemit tne passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube, inspection.
c.
The tubes selected as the second and third samples (if required by Table 4.4-T.) during each inservice inspection may be subjected to a partial tube inspection provi;ed:
1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes witn imperfections were previously found.
2.,
The inspections include those portions of the tubes where imperfections were previously found.
The results of each : ample inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 Cne 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 cegraded tubes.
C-3 More than 10% of the total tubes inspectad are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previcusly degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.
l TROJAN-UNIT 1 3/4 4-7 Amendment tio. 61
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of steam generator tunes shall be performed at the following frequencies:
a.
The first inservice inspection snall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections snall be performed at irtervals of n:t less than 12 nor mere than 24 calendar months af ter the previous inspection.
If two consecu-tive inspections following service under AYT conditions, not including the preservice inspection, result in all inspection results falling into tre C-1 category or if two consecutive inspections demonstrater 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.
b.
If the results of the inservice inspection of a steam gen-erator conducted in accordance with Table 4.4-2 at 40 montn intervals fall in Category C-3, the inspection frequency snall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequ1nt inspections satisfy the criteria of Specifica-tion 4.4.5.3.a; the interval aay then be extended to a maximum of onc. per 40 months.
I c.
Acditional, unscheduled inservice 'nspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 curing the shutdown subsequent to any of the following conditions:
1.
Primary-to-secondary tubes leaks (not including leaks originating frm; *"be-to-tube sheet welds) in excess of the limits of c;eci 'ication 3.4.6.2.
2.
A seismic occurrence s' eater than the Operating Basis Earthquake.
3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam,line or feedwater line break.
TRGJAN-UNIT 1 3/4 4-8 Amencment i;o. 61
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued).
4.4.5.4 Acceptance Criteria a.
As used in the Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing incications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring en either inside or
outside of a tube.
3.
Degraded Tube means a tube containing imperfections >20%
of tne nominal wall thickness caused by degradation.
4.
% Degradation means the percentage of the tube wall tnickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceecs the plugging limit.
A, tube containing a defect l
1s defective.
6.
Plugging Limit means the imperfection cepth at or beyond wnicn tne tuoe shall be removed from service because it may become unserviceable prior to the next inspection and is equal to (40)% of the nominal tube wall thickness.
7.
Unserviceable describes the conaition of a tube if it ieaks 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 as specified in 4.4.5.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tuce from tne point of entry (hot leg side) ccmpletely around the U-bend to the top support of the cold leg.
TROJ AN-UNIT 1 3/4 4-9 Amendment No. 61
l REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 9.
Preservice Inspection means an inspection of the full length of eacn tuce in eacn steam generator performed by eddy current' techniques prior to service establish a caseline condition of the tubing.
This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reports a.
Following e3ch inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
b.
The complete results,cf the steam generator tube inservice inspection shall be reported on an annual basis for the perioc l
l in which the inspection was completed.
This report shall l
include:
1.
Number and extent of tubes inspected.
i 2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
l 3.
Identification of tubes plugged.
l c.
Results of steam generator +wbe inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to l
resumption of plant operation.
The written followup of this l
report shall provide a description of investigations conducted to determine cause of the tube cegradation and corrective measures taken to prevent recurrence.
s 1
TROJ AN-UNIT 1 3/4 t.-9a Amendment No. 61
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1. 2 and 3 com-ponents shall be maintained in accordance with Specification 4.4.10.1.
APPLICABILITY:
ALL MODES ACTION:
a.
With the structural integrity of any ASME Code Class 1 com,.
ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations, or comply with ACTION item d below.
b.
With the structural integrity of any ASME Code Class 2 com-ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit.or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200'F, or comply with ACTION item d below.
c.
With the structural integrity of any ASME Code Class 3 com-ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service, or comply with ACTION item d below.
d.
If the requirements of the above ACTION items are not met, in lieu of the requirements of Specification 3.0.3, an evaluation can be performed to determine the consequences of continuing to operate with reduced structural integrity or with a temporary repair made to the affected component (s),
or be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, e.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10.1 ASME Code Class 1, 2, and 3 components shall be inspected in accordance wit!. the requirements of Specification 4.0.5.
In addition, eacn reactor coolant pump flywneel shall be inspected per the recommenca-tions of Regulatory Position C.4.b of Regulatory Guide 1.14, Revisicn 1, August 1975.
il TROJ AN-UNIT 1 3/4 4-29 Amendment No. 61
APPLICABILITY BASES 4.0.5 This specificatinn ensures that inservice inspection of ASME Code Class 1, 2 aad 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a.
Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Tecnnical Specifications.
This specification includes a clarification of the frequencies for performing tha inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in sur-veillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing th'e required inservice inspection and testing activities.
Under the terms of this specification, the more restrictive require-ments of the Technical Specifications take precedence over the ASME Boiler and Pressure Yessel Code and applicable Addenda.
For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an CPERATIONAL MODE or other specified applicabil.ity condition 'akes precedence over the ASME Boiler and Pressure Vessel Code provision-which allows pumps to be tested up to one week after return to normal operation. And for example. the Technical Specification defini-tion of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its speci-fied function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable.
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4 TROJ AN-UNIT 1 B 3/4 0- 5 Amendment "o.
51
r REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS CONTINUED system and the secondary coolant system (primary-to-seconoary leakage =
500 gallons per day per steam generator).
Cracks having a primary-to-secondary leakage less than this limit curing operation will have an adequate margin of safety to withstand the loacs imposed during ncrmal operation and by postulated accidents.
Operating plants have cemonstrated l
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 shutcown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with the all volatile treatment ( AVT) of secondary coolant. However, even if a defect of similar type shculd develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition i
of Specification 4.4.5.4.a is 40*. of the tube ncminal wall thickness.
Steam generator tube inspections of cperating plants have demonstrated the capability to reliably detect degradation that has penetrated 20*. of the l
original tube wall thickness.
l l
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 Wy-current inspection, and revision l
of the Technical Specifications, f necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l
3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Raactor Coolant Pressure Soundary.
These detection systems are consistent with the recommencations of Regulatory Guide 1.45, " Reactor Ccolant Pressure Boundary Leakage Detection Systems," May 1973.
i TROJAN-UNIT 1 B 3/4 t-2a Amendment No. 61
0 s
REACTOR COOLANT SYSTEM BASES The actual shift in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irraciation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of tne reactor vessel. The heatup and cooldown cur /es must be recalculatec when the ARTNOT determined from the surveillance capsule is different from the calculated aRTNOT for the equivalent capsule radiation exposure.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens ano the frequencies for removing and testing these. specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 1
The limitaticns imposed on pressurizer.heatup and cooldown and spray water temperature differential are provided to assure that the
. pressurizer is operated within the cesign criteria assumed for the
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fatigue analysis performed in accordance with the ASME Code requirements.
3/4.4.10 STT,UCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 componc.1ts ensure that the structural integrity and operational readiness of these components will be maintained at. an acceptable level r
l throughout the life of the plant.
These programs are in accordance with e
l Section XI of the ASME Boiler and Pressure Vessel Code and applicable l
Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).
l Components of the reactor coolant system were designed to provice access to permit inservice inspections in accordance with Section XI of l
the ASME Boiler and Pressure Yessel Code,1971 Ecition and Adcenda I
through winter 1972.
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ii TACJ AN-UNIT 1 B 3/4 4-9 Amen: ment No. 61 l