ML20199F784
| ML20199F784 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/19/1999 |
| From: | Nolan M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20199F790 | List: |
| References | |
| NUDOCS 9901220060 | |
| Download: ML20199F784 (11) | |
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'4 UNITED STATES l
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NUCLEAR REGULATORY COMMISSION e
~f WASHINGTON, D.C. 30666-0001
%,.....f ENTERGY OPERATIONS. INC.
DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.199 License No. NPF-6 i
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated June 29,1998, 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 l
Act, and the rules and regulations of the Commission; 1
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; 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.
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l 9901220060 990119
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PDR ADOCK 05000368 P
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2.
Accordingly, the license 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-6 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 199, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance with full implementation within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION h.' h Y..u p c $ A M. Christopher Nolan, Project Manger Project Directorate IV-1 Division of Reactor Projects lll/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: January 19, 1999 s
= -. - - -..
l' ATTACHMENT TO LICENSE AMENp_ MENT NO 199 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 i
I Revise the following 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.
l REMOVE PAGES INSERT PAQgS vi vi j
XI XI*
Xil XII 3/4 4-3 3/4 4-3 3/4 4-28 3/4 4-28 l
B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2*
i B 3/4 4-12 B 3/4 4-12 i
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- Overleaf pages provided to maintain document completeness. No changes contained on these pages.
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INDEX e
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION...........
3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOWN................................
3/4 4-3 3/4.4.3 SAFETY VALVES - OPERATING...............................
3/4 4-4 3/4.4.4 PRESSURIZER.............................................
3/4 4-5 3/4.4.5 STEAM GENERATORS........................................
3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...............................
3/4 4-13 Reactor Coolant System Leakage..........................
3/4 4-14 3/4.4.7 CHEMISTRY...............................................
3/4 4-15 3/4.4.8 SPECIFIC ACTILITY.......................................
3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System..................................
3/4 4-22 Pressurizer.............................................
3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components...................
3/4 4-26 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................
3/4 4-27 3/4.4.12 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM...
3/4 4-28 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS..................................
3/4 5-1 l
ARKANSAS - UNIT 2 VI Amendment No. M,w,e,4M,199 g
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o
.INDEx d
BASES 4
SECTION PAGE a
I 3/4.0 AP P LI CA3 f f,.'g..............................................
B 3/4 0-1 3/4.1 "REACTIVIT't CONTROL SYSTEMS i
3/4.1.1 SORATION CONTR0L........................................
3 3/4 1-1, i
3/4.1.2 BORTEION SYSTINS........................................
S 3/4 1-2 4
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3/4.1.3 MDVARLE CONTROL ASSIH5 LIES..............................
,B 3/4 1-3 3/4.2 POWZR DISTRIBUTION LIMITS 1
3/4.2.1 LINEAR IFAT RATE........................................
B 3/4 2-1 i
j 3/4.2.2 RADIAL PEAKING FACT 0RS..................................
B 3/4 2-2
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3/4.2.3 AIIMUTEAL POWER TILT....................................
B 3/4 2-2 l
3/4.2.4 DNSR MARGIN...........
........n.......................
,5 3/4 2 3 j
'3/4.2.5 RCS FLON 3 ATE..........;................................
3 3/4 2-4 2n.2..
mExeTOR c003 ANT c03.D LEG TuxrERuvRE....................
. 3n 2-4 3/4.2.7 AXIAL SHAPE 2NDEX.......................................
3 3/4 2-4
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3/4.2.8 PRESSURIZER PRESSURE....................................
S 3/4 2-4 3/4.3 TNSTRUMENTATION f
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3/4.3.1 PROTECTIVE INSTRUMENTATION..............................
2 3/4 3-1 4
l 3/4.3.2 ENGINEERED 311ETY TEATURE INSTRUMENTATION................
B 3/4 3-1 f
3/4.3.3 MONITORING INSTRUMENTATION....,..........................
5 3/4 3-2 1
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i ARKANSAS - UNIT 2 XI Amendment No. e4,aa,40,191 JUL 3 E
W INDEX RASES SECTION PAGE 3/4.4 RFACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOIANT CIRCUIATION...........
D 3/4 4-1 3/4.4.2 and 3/4.4.3 SArETY VALVES.................................
B 3/4 4-1 3/4.4.4 PRESSURIZER.............................................
B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................
B 3/4 4-2 3/4.4.6 REACTOR COOIANT SYSTEM LEAKAGE..........................
B 3/4 4-3 3/4.4.7 CHEMISTRY...............................................
B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.......................................
B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................
B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY....................................
B 3/4 4-11 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................
B 3/4 4-11 3/4.4.12 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM..........
B 3/4 4-12 l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS..................................
B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..............................
B 3/4 5-1 3/4.5.4 RETUELING WATER TANK (RWT)..............................
B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.....................................
B 3/4 6-1 3/4.6.2 DEPRESSURIZATION, COOLING, AND pH CONTROL SYSTEMS.......
B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES............................
B 3/4 6-6 3/4.6.4 COMBUSTIBLE GAS CONTROL.................................
B 3/4 6-6 a
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ARKANSAS - UNIT 2 XII Amendment No. M,M,M,M4,199
REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIA i 3%*.
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APPLICABILITY:
MODE 4 with Tc > 220'F.
l ACTION:
With no pressurizer code safety valve OPERABLE, reduce Tc to 5 220'r within l
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
If found outside of a i 1% tolerance band, the setting shall be adjusted to within i 1% of the lift setting shown.
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ARKANSAS - UNIT 2 3/4 4-3 Amendment No. M9,MO,W,199 i
9 REACTOR COOLANT SYSTEM LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM l
LIMITING CONDITION FOR OPERATION 3.4.12 The LTOP system shall be OPERABLE with each SIT isolated that is pressurized to 2 300 psig, and:
a.
Two LTOP relief valves with a lift setting of 5 430 psig, or I
b.
The Reactor Coolant System depressurized with an RCS vent path l
2 6.38 square inches.
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APPLICABILITY: MODE 4 with Tc 1220'F, MODE 5, MODE 6 with reactor vessel head in place.
I ACTION:
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a.
With one LTOP relief valve inoperable in MODE 4, restore the inoperable valve to OPERABLE status within 7 days or depressurize and vent the RCS through a 2 6.38 square inch vent path within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With cne LTOP relief valve inoperable in MODE 5 or 6, restore j
the inoperable relief valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or depressurize and vent the RCS through a 2 6.38 square inch vent path within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
c.
With both LTOP relief valves inoperable, depressurize and vent the RCS through a 2 6.38 square inch vent path within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, l
d.
With a SIT not isolated and pressurized to 2 300 psig, isolate the affected SIT within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
If the affected SIT is not isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, either:
(1) Depressurito the SIT to < 300 psig within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or l
(2) Increase cold leg temperature to > 220*F within the next 12 l
hours.
e.
The provisions of Specification 3.0.4 are not applicable.
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4 ARKANSAS - UNIT 2 3/4 4-28 Amendment No. 444,199 l.
3/4.4 REACTOR COOLANT SYSTEM
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BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above the limits specified by Specification 3.2.4 during all normal operations and anticipated transients.
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for renoving decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for ramoving 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.
j The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.
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 valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
Two safety valves are required in MODES 1, 2 and 3.
One safety valve is required in MODE 4 with T, > 220*F.
For the remainder of MODES 4, 5 and 6 with the reactor vessel head in place, everpressure protection is provided by the operating procedures and LCO 3.4.12, " Low Temperature Ovezpressure Protection (LTOP) System".
During operation, all pressurizer cocie 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 Reactor Coolant System pressure to within its safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the steam dump valves.
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I ARKANSAS - UNIT 2 B 3/4 4-1 Amendment No. H,n,H9,199
ArActoR cootANT SYSTEM RASES Demonstration of the safety valves' lift setting will occur only during shutdown and will be perfonned in accordance with the provisions of Section XI of the AsME Soiler and pressure vessel Code,_
j 3/4.4.4 TRESSURIEER A steam bubble in the pressuriser ensures that the acs is not a hydraulically solid system and is onpable of accomunodating pressure surges l
during operation. The steam bubble also protects the pressuriser code i
safety valves against water relief. The steam bubble functions te salieve l
Rcs pressure during all design transients.
j The requirement that 150 RN of pressuriser heaters and their associated controls be capable of being supplied electrical power from an j
emergency bus provides assurance that these heaters can be emergized during
-a loss-of-offsite power condition to maintain astural circulation at rot STAND 8Y.-
j 3/4.4.5 STEAM GDrERATons The surveillance Requirements Tor taspection 32 the steam generator i
tubes ensure that, the structural integrity of this portion of the Rc8 will i
be maintained. The program for inservice inspection of steam generator
{
tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
j Inservice inspection of steam generator tubing is essential in order to j
-maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to j
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.
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The plant is expected to be ope.tated in a manner such that the j
secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the a
-secondary coolant chemistry is not maintained within these limits, j
localised sorrosion may likely-result.in-stress corrosion.szacking..
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary. coolant system and the secondary coolant system (primary-to-secondary leakage
= 150 gallons per day per steam generator). Cracks having a primary-to-j
-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 150 gallons per day per steam generator saa readily be detected by radiation monitors en the secondary system.
-1.sakage in excess of this limit will require plant shutdown and an unscheduled laspection, during which the leaking tubes will be located and plugged or sepaired.
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ARKANSAS ~1FNIT 2 5 3/4 4-2 Amen h t pe. 40,664*184
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l REACTOR CCOZAMT SYSTEM RASE 3 3/4.4.12 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM l
Low temperature overpressure protection (LTOP) of the RCS, including the reactor vessel, is provided by redundant relief valves on the pressurizer which discharge from a single discharge header.
Each LTOP relief valve is a direct action, spring-loaded relief valve, with orifice area of 6.38 in* and a lift setting of 5 430 psig, and is capable of mitigating the worst-case design basis pressurization event. The LTOP design basis event is a simultaneous injection of two HPSI pumps and all
. three charging pumps to the water-solid RCS.
The analyses assume that the safety injection tanks (SITS) are either isolated or depressurized such that they are unable challenge the LTOP relief setpoints.
t Since neither the LTOP reliefs nor the RCS vent is analyzed for the pressure transient produced from SIT injection, the LCO requires each SIT j
that is pressurized to 2 300 psig to be isolated. The isolated SITS must have their discharge valves closed and the associated MOV power supply breakar in the open position. The individual SITS may be unisolated when i
pressurized to < 300 psig. The associated instrumentation uncertainty is not included in the 300 psig value and therefore, the procedural value for unisolating the SITS with the LTOPs in service will be reduced.
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The LTOP system, in combination with the RCS heatup and cooldown i
limitations of LCO 3.4.9.1 and administrative restrictions on RCP l
operation, provides assurance that the reactor vessel non-ductile fracture limits are not exceeded during the design basis event at low RCS temperatures. These non-ductile fracture limits are identified as LTOP pressure-temperature (P-T) limits, which were specifically developed to provide a basis for the LTOP system. These LTOP P-T limits, along with the LTOP enable temperature, were developed using guidance provided in ASME Code Section XI, Division 1, Code Case N-514 that mandates that *LTOP systems shall limit the maximwn pressure in the vessel to 110% of the pressure determined to satisfy Appendix G, paragraph G-2215 of Section XI, Division 1".
The enable temperature of the LTOP isolation valves is based on any RCS cold leg temperature reaching 220'F (including a 20'F uncertainty).
Although each relief valve is capable of mitigating the design basis LTOP event, both LTOP relief valves are required to be OPERABLE below the enable l
temperature to meet the single failure criterion of NRC Branch Technical Position RSB 5-2, unless any RCS vent path of 6.38 ina (equivalent relief l
valve orifice area) or larger is maintained.
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i ARKANSAS - UNIT 2 5 3/4 4-12 Amendment No. 440,199
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