ML20080F367
| ML20080F367 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/05/1995 |
| From: | Kalman G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080F370 | List: |
| References | |
| NPF-06-A-158 NUDOCS 9501200214 | |
| Download: ML20080F367 (10) | |
Text
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j' UNITED STATES o
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- ,E NUCLEAR REGULATORY COMMISSION o
WASHINGTON D.C. 20E56 ENTERGY OPERATIONS INC.
DOCKET NO. 50-368 i
ARKANSAS NUCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 158 License No. NPF-6 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated November 29, 1994, as supplemented by letters dated December 20 and 21, 1994, 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, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license 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 requirements have been satisfied.
9501200214 950105 PDR ADOCK 05000360 P
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2.
Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-6 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.158, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
t 3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION a
nior P Project Directorate IV-1 i
Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications i
Date of Issuance:
January 5, 1995 I
l l
ATTACHMENT TO LICENSE AMENDMENT NO. 158 FACILITY OPERATING LICENSE NO. NPF-6 1
DOCKET NO. 50-368 Replace the following pages of the Appendix "A" Technical Specifications with i
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 4-12 3/4 4-12 B3/4 4-2 B3/4 4-2
(
I
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR 0PERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of < 910 cubic feet (equivalent to 182% of wide range indicated level) and both g
pressurizer proportional heater groups shall be OPERABLE.
I APPLICABILITY: MODES 1. 2 and 3.
ACTION:
(a) With the pressurizer inoperable due to water volume >910 cubic feet.
l be in at least HOT SHUTDOWN with the reactor trip bre' kers open a
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(b) With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters, 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 HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer sater 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 groups shall be determined to be OPERABLE:
(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 emergency power is,
available to the heater groups, and (b) At least once per 18 months by verifying that the summed power i
consumption of the two proportional heater groups is > 150 KW, i
ARKANSAS - UNIT 2 3/4 4-5 Amendment No. 20 1
m-a 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 OPERABLE status prior to increasing T above 200'r.
avg SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
Note: The surveillance requirements of Specification 3.4.5 do not apply to the special steam generator tube inspection to be performed during the 2P95-1 outage scheduled to begin on January 6, 1995. The scope and expansion criteria for this inspection are specified in correspondence to the NRC submitted under separate cover. The scope and criteria shall be approved by the NRC prior to exiting Mode 5.
The results of this inspection shall be reviewed by the Plant Safety Committee prior to resumption of plant operation and reported to the NRC within 30 days of resumptien of plant operation.
4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam 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 classification, 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 steam generators; the tubes selected for these 1.Spections shall be selected on a random basis excepts Where experience in similar plants with similar water chemistry a.
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:
ARKANSAS - UNIT 2 3/4 4-6 Amendment No. 158
.=
REACTOR Coo 1 ANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1.
All nonplugged tubes that previously had detectable wall penetrations (>206).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to specification 4.4.5.4.a.9) l 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 be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
The tubes selected as the second and third samples (if required c.
by Table 4.4-21 during each inservice inspection may be subjected to a partial tube inspection provided:
1.
The tubes selected for these samples include 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 it 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 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 significant (>106 ) further wall penetrations to be included in the above percentage calculations.
ARKANSAS - UNIT 2 3/4 4-7 Amendment No. 158
e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5,3 Inspection Frequencies - The above required inservice inspections of steam generater tubes shall be perfomed at the following frequencies:
The first inservice inspection shall be performed after 6 a.
Effective Full Power Months but within 24 calendar 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 service under AVT conditions, not including the preservice inspection, result in all inspec-tion results falling into the C-1 category or if two consecu-tive 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 Table 4.4-2 at 40 month 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.a; the interval may then be extended to a maximum of once per 40 months.
Additional, unscheduled inservice inspections shall be performed c.
on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 durin quent to any of the following conditions:g the shutdown subse-1.
Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating Basis Earthouake.
3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam line or feedwater line break.
ARKANSAS - UNIT 2 3/4 4-8
e TABIE 4.4-2 STEAM GENERATOR TUBE INSPECTION l
IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMP1.E INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A 11/A l
5 Tubes per 5.G.
C-2 Plug or sleeve def ec-C-1 None N/A N/A tive tubes and inspect additional 25 tubes in Plug or sleeve defec-C-1 None this S G.
C-2 tive tubes and inspect C-2 Plug or sleeve additional 45 tubes in defective tubes this S.G.
Perform action for C-3 C-3 result of first sample Perforta action for C-3 C-? result of first N/A N/A sample C-3 Inspect all tubes in All other this S.G., plug or S.G.s are None N/A N/A sleeve defective tubes C-1 and inspect 25 tubes in each other S.G.
Some S.G.s Perform action for A/A N/A C-2 but no C-2 result of second additional sample Special Report S.G.
are to NRC per C-3 Specification 6.9.2 Additional Inspect all tubes in S.G.
is C-3 each S.G. and plug or sleeve defective tubes.
Special N/A N/A Report to NRC per Spec. 6.9.2.
S = 3 N,% Where N is the number of steam generators in the unit, and n is the number of steam generators inspected D
during an inspection 4
ARKANSAS - UNIT 2 3/4 4-12 Amendment No. M,ua,Ma,158
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3/4.4. erAr* Tom Erwirlwr sysTrw o
RAMA i
3/4.4.1 mEACT0if COntlwT fimp3 AND EnnflMT Cf tetfLATitW t
i The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumpe in operation, and maintain DNBR above the limits specified by Specification 3.2.4 during all normal operations and l
anticipated transients.
In MODE 3, a single reactor coolant loop provides sufficient heat removal capabi!!ty for removing decay heat; however, single failure considerat?ons require that two loops be OPERABLE.
In MODEd 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to be OPERABLE.
The operation of one Reactor Coolant pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and l
produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with l
boron reductions will, therefore, be within the capability of operator recognition and control.
l I
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 420,000 lbs. per hour of saturated steam at the i
valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient to limit the Re.ncver Coolant System pressure to o tthin its Safety Limit of 2750 psia follsving a complete loss of turbine generator load while operating at RATED TIERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., na credit is taken for a direct reactor trip on the loss of 1
turbine) and also assuming no operation of the steam dump valves.
ARKANSAS - UNIT 2 8 3/4 4-1 Amendment No. 25,49-149 l
l i
- i REACTOR COOLANT SYSTEM RASES Demonstration of the safety valves' lift setting will occur only l
during shutdown and will be performed in accordance with the provisions of section XI of the ASME Boiler and Pressure Vessel
- Code, r
3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code j
safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.
The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss-of-offsite power condition to naintain natural circulation at HOT STANDBY.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will r
be maintained. The program for inservice inspection of steam generator l
tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
!aservice inspection of steam generator tubing is essential in order to i
maintain surveillance of the conditions of the tubes in the event that l
there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the seccndary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not raintained within these lindts, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the lindtation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage
= 0.5 GPM per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate nargin of safety to withstand the loads imposed during normal operation i
and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 0.5 GPM 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 leakang tubes will be located and plugged or repaired.
ARKANSAS - UNIT 2 B3/4 4-2 Amendment No. 34,158 I
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