ML20198G003
| ML20198G003 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/23/1997 |
| From: | Kalman G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20198G010 | List: |
| References | |
| NPF-06-A-187 NUDOCS 9801120228 | |
| Download: ML20198G003 (11) | |
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,j NUCLEAR REGULATORY COMMIS810N 2
W ASHINGTON, D.C. 3006H001
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ENTERGY OPEP.ATIOG. INC'.
DOCKET NO. 50-368 ARKANSAS l.dCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.187 i
License No. NPF-6 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated April 24, 1997, complies with the standards and Act), and the Comission's raies and regulations set forth(the requirt.ments of tie Atomic Energy Act of 1954, as amended in 10 CFR Chapter I; B.
The facility will operate in conformity with the appitcation, the provisions of the Act, and the rules and regu'ations of the Commission; C.
There is reasonable assurance:
(i) 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 Commission's regulations; 0.
The issuance of this license e,mendment 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 requirements have been satisfied.
9001120228 971223 PDR ADOCK 05000368 P
2-i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license ah,endment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-6 is hereby amended to read as follows:
2.
Technical Snecifications The Technical Specifications contained in Appendix A, as revised thrcugh Amendment No.187, are hereby incorporated in the license.
The licensee shall optrate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective within 30 days of the issuance date.
FOR THE NUCLEAR REGULATORY COMMISSION
/
1:
George Ka an. Senior Project Manager Project Directorate IV-1 Division of Reactor Projects !!!/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: Decea6er 23, 1997
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4 ATTACHMENT TO LICENSE AMENDMENT NO.187 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50 368 Revise the vollowing pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE PAGES INSERT PAGES 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 4-10 3/4 4-10 3/4 4-12 3/4 4-12 3/4 4-12a 3/4 4-12a B 3/4 4-3 8 3/4 4-3
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ZRf15URIltt LIMITING CONDITION POR' CPERATION 3.4.4 The presseriter shall be OPEAABLE with a enter volume of = 910 pressortaer(proportional h~ eater groups shall be OPERASLE.equivalant to <
a cubic feet E
APPLICA3fLITY: sc0ts 1. 2 and J.
Mt (a) With the pressurher Jnoperable due to water volume 1910 cubic feet.
l be in at least NOT SHUTDOWN with the reactor trip breakers open within 12 hows.
(b) iUth the pressuriser inocerable due to an inoperable emergency power supply to the pressuriser betters, either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least NOT SHllTDOWN i
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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$URVE!LLANC[ RIQUIREMENTS 4.f.4.1 The pressurf ter water volume shall be determined to be'within its l
limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.4.2 The pressurizer proportional heater yovps shall be determined i: be OPIUJLI:
(a) At least once per.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying smergency power is,
available to the heater groups., and (b) At least once per 18 months by verifying that the summed power consvaption of the two proportional heater proups is y,,150 MW.
ARKANSAS - DNIT 2 J/4 4-5 Amendment No. 20 f'
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STEAM GENERATORS LIKITING COND1110N POR FFERATION 3.4.5 Each steam generator shall be OPERABLE.
j APPLICA81LTTY: MODES 1,2, 3 and 4.
ACTION:
With one or more steam generators inoperabla, restore the inoperable generator (e) to OPERABLE status prior to increasing Tavg above 200'r. '
SURVEILLANCE RECTUIRDENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice fnopection crogram and the requirements of Specification 4.0.5.
I 4.4.5.1 steam Ganerator sample selection and Inspection - Each steam generator shall be deteruined DFERABLE 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 iube sample selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as spe:1flad in Tables 4.4-2 and 4.4-3.
The inservice inspection or steam generator tubes shall be performed at the frequenci6s specified in specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of specification 4.4.L.4.
The tu>es selected for each inservice inspection shall include at least 36 of the total number of tubes in 211 steam generators and at least 206 of each type of installed sleeves; the tubes selected for these inspartions shall be selected on a random basis excepts a.
Where 2xperience in similac plants with similar water chamistry 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 the preservice, inspection) of each steam generator shall include:
4 ARKANSAS - UNIT 2 3/4 4-6 Amendment No.M4,187 w
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REACTOR CooLAN7 system SURVEILLANCE REQUIREMENTS (ContinuOd) 1.
All monplugged tubes that previously had detectable well penetrations (>204).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to specification 4.4.5.4.a.9) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall se recorded and an adjacent tube shall be selected and subjected to a tube it spectiJn.
c.
The tubes selected as the second and third samples (if required by Tables 4.4-2 or 4.4-3) during each inservice inspection may be subjected to a partial inspection provided:
1.
The tubes selected for these samples incluse the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections were previously found.
The result of each sample inspection shall be classified into one to 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 One or more tubes, but not more than 1% of the total tub 6s inspected are defective, or between St and 10% of the total tubes inspected are degraded tubes.
C-3 Nore than 10% of the total tubes inspected are degraded tubes or more than 16 of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit rignificant (>106) further wall penetrations to be included in the above percentage calculations.
ARKANSAS - UNIT 2 3/4 4-7 Amenchment No.Me,187 w
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REACTOR C001 ANT SYSTD4 SURVE!L1ANCE REQUIREMDfTs (Continued) 4.4.5.3 Inspection Freeuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 salendar months of initial criticality.
Subsequent inservice 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 servios under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 estogory or if two consecutive inspections 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.
b.
If the results of the inservice inspection of a steam generator conducted in accordance with Tables '4.4-2 and 4.4-3 at 40 month l
intervals fall into 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 4.4.5.3.as the interval may then be extended to a maximum of once per 40 months.
c.
Additional, unscher'Aed inservice inspections shall be performed on each steam genecator in accordance with the first sample inspection specified in Tables 4.4-2 and 4.4-3 d. iring the shutdown l
subsequent to any of the following conditions:
1.
Primary-to-secondary tube leaks (not including leaks l
originating from tube-to-tube sheet welds) in excess of the limits of specifiettion 3.4.6.2.
i 2.
A seismic occurrence greater than the operating Basis Earthquake.
l 3.
A loss-of coolant accident requiring actuation of the l
engineered safuguards.
4.
A main steam line or feedwater line break.
l ARKANSAS - UNIT 2 3/4 4-8 Amendment No.187
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REACTOR C00LANT SYSTD4 SURVEILLANCE REQUIRD(ENTS (Continued) 10.
Preservice Inspection means an inspection of the full length l
ei each tube in each steam generator performed by oddy current techniques prior to service to establish a baseline 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 or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Tables 4.4-2 and 4.4-3.
Defective tubes may be l
repaired in accordance with:
1)
B&W Topical Report RAN-2045PA-00 as supplemented by the information provided in Bow Report 51-1212539-00, "BWNS Kinetic sleeve Design-Application to ANO Unit 2".
2)
CENO Report CEN-630-P, " Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight sleeves," Revision 01, dated November 1996. The post weld heat treatment describee in CEN-630-P shall be performed.
4.4.5.5 Reports s.
Following each inservice inspection of steam generator tubes the number of tubes plugged or sleeved in each steam gnnerator 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 shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of well-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged or sleeved, c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a tpecial Poport pursuant to specification 6.9.2 as denoted by Tanies 4.4-2 and 4.4-3.
l Notification of the Conadssion will be made prior to resumption of plant operation. The written special Report shall provide a description of investigations conducted to deterndne cause of the tuke degradation and cor:ective measures taken to prevent recurrence.
ARKANSAS - UNXT 2
.3/4 4-10 Amendment No. M,m,W,M4,187
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TABLE 4.4-2 STEAM GENERATOR NON-REFAIRED TUSE INSFECTION l
IST SAMFLE INSFECTION 2ND SAMPLE INSFECTION 3RD SAMPLE INSFECa L' I
Sample size Aesult Action W ired Result Action % ired Result Action % g;1 red A minianam of C-1 None N/A N/A N/A N/A D Tubes per 0.0.
C-2 Flug or sleeve defec-C-1 None N/A I
N/A tive tubes and inspect additior4 25 tubes in Flug or sleeve defec-C-1 mone this S.G.
C-2 tive tubes and inspect C-2 Flug or sleeve additional 45 tubes in defective tubee l
this 3.G.
l Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample l
C-3 Inspect all tubes in Other S.G.
this 3.G., plug or is C-1 Noae N/A N/A l
sleeve defective tubes and inspect 25 tubes in the other S.G.
Other S.G.
Perform action for N/A N/A is C-2 C-2 result of second sanple Special Neport to NRC per Specification 6.9.2 Other 3.G.
Inspect all tubes in is C-3 the other S.G. and plug or sleeve defective tubes.
N/A N/A Special Neport to NRC per Spec. E.9.2 l
D = 3 2, & Where n is the number of stessi generators inspected during an inspection.
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-t No. M,4aa,444,4sa)87 ARKANSAS - UNIT 2 3/4 4-12
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TABLE 4.4-3 STEAM GENERATOR TUSE SLEEVE INSFECTION IST SAMPLE INSPECTION 2ND SAMPLE INSFECTION 2 7 w Size mesult Action Required nesult Action = ;; ired 1
4 Maimum of 20% of C-1 None N/A N/A each type of installed sleeve per S.G.
(1)
C-2 Flug tubes contalning C-1 None defective sleeves and inspect all remaining installed sleeves in this t
s.G.
C-2 Flug tubes containing defective sleeves C-3 Forform action for C-3 result of first sample C-3 Inspect all installed Other 3.G. is None sleeves in this t.G.,
C-1 plug tubes containing defective sleeves and inspect 100% of the installed sleeves in the other 3.G.
Special Report to NRC per Other S.G. is Flug tubes containing defective sleeves Specification 6.9.2 C-2 Other 5.G. is Inspect all installed aleewee in each *.G.
C-3 and plug tubes containing defective sleeves special poport to NRC per spec. 5.9.2 *
(1) Each sleeve type is considered a separate population for 4 termination of easyle expenslom.
2 ARKANSAS - UNIT 2 3/4 4-12a W t No.187
0 REACTOR CLOLANT SYSTEM eases Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tubes esaminations.
Plugging or sleeving will be required for all tubes with imperfections,
exceeding the plugging or repair limit as defined in surveillance Requirement 4.4.5.4.a.
Defective tubes may be repaired by sleeving in accordance with the B&W Topical Report EAN-0045PA-00 as supplemented by the information provided in B&W Report 51-12125!9-00, "BWN3 Kinetic Sleeve Design-Application to ANO Unit 2" or CENO Report CEN-630-P, 'P.epair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight Sleeves," Revision 01, dated November 1996. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the tube wall thickness.
For sleeved tubes, the adequacy of the systen that is used for periodic inservice inspection will be validated. Additionally, upgraded testing methods will be evaluated and appropriately implemented as better methods are developed and validated for commercial use.
Whenever the results of any steam generator tubing inservice inspection fell into Category C-3 certain results will be reported in a special Report to the commission pursuant to specification 6.9.2 as denoted by Table 4.2-2 or 4.4-3.
l Notification of the Commission will be made 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, laboratery examinations, tests. additional eddy-current inspection, and revision of the Technical specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systema required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendat'.ons of Regulatory Guide 1.45, "Resctor Coolant Pressure Boundary Leakage Detection Systems" May 1973.
3/4.4.6.2 REACTOR COOLANT SYSTEM LEAXAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPH. This threshold value is sufficiently low to ensure early detection of additional leakage.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systams.
The surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve f ailure and consequent intersystem LOCA.
Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be conJidered as a portion of the allowed limit.
ARKANSAS - UNIT 2 3 3/4 4-3 Amendment No. 94,44b 444,Gedee d:::d t/00/01,444,181
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