ML20211P949
| ML20211P949 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/09/1999 |
| From: | Gramm R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211P953 | List: |
| References | |
| NUDOCS 9909140144 | |
| Download: ML20211P949 (10) | |
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t UNITED STATES j
NUCLEAR REGULATORY COMMISSION 7
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WASHINGTON, D.C. 20566-0001 4 *****,o 9
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ENTERGY OPERATIONS INC.
DOCKET NO. 50 313 l
l ARKANSAS NUCLEAR ONE. UNIT NO.1 l
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No.199 License No. DPR-51 j
l 1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated l
April 9,1999, as supplemented by "0CAN079901, Application for Amends to Licenses DPR-51 & NPF-6 to Change TS & Subsequent Relief Request from Post Accident Sampling Requirements of NUREG-0737.TS Administrative Control requirements,[[NUREG" contains a listed "[" character as part of the property label and has therefore been classified as invalid. & RG 1.97,rev 3,affected|letter dated July 14,1999]], 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 1:
l B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; l
C.
There is reasonable assurance (i) that the activities authorized by this l
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 regulatiens; l
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 l
L 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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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to th.s license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-51 is hereby amended to read as follows:
- 2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through j
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 and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A /$n-Robert A. Gramm, Chief, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 9, 1999
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ATTACHMENT TO LICENSE AMENDMENT NO.199 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313
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f Replace the following pages of the Appendix A Technical Specifications with the attached.
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l revised pages.. The revised pages are identified by amendment number and contain marginal I
lines indicating the areas of change.
l Remove Insert ii ii iv iv 54 54-55 55 80 80 81 82 83-84 85 85a 85b 85c
'85d 86 87 88 89~
90 91
'146 146 146a 146a I
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SECTION TITLE PAGE 3.14 HYDROGEN RECOMBINERS 66e 3.15 FUEL HANDLING AREA VENTILATION SYSTEM 66g 3.16 SHOCK SUPPRESSORS (SNUBBERS) 661 3.17 FIRE SUPPRESSION WATER SYSTEM 66m 3.18 FIRE SUPPRESSION SPRINKLER SYSTEMS 66n 3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 66o 3.20 FIRE HOSE STATIONS 66p 3.21 FIRE BARRIERS 66q 3.22 REACTOR BUILDING PURGE FILTRATION SYSTEM 66r 3.23 REACTOR BUILDING PURGE VALVES 66t 3.24 EXPLOSIVE GAS MIXTURE-66u 3.25 RADIOACTIVE EFFLUENTS 66v 3.25.1
, Radioactive Liquid Holdup Tanks 66v 3.25.2 Radioactive Gas Storage Tanks 66w 4.
SURVEILLANCE REQUIREMENTS 67
'4.1 OPERATIONAL SAFETY ITEMS 67
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4.2 REACTOR COOLANT SYSTEM SURVEILLANCE 76 j
4.3 TESTING FOLLOWING OPENING OF SYSTEM 78 4.4 REACTOR BUILDING 79 4.4.1 Reactor Building Leakage Tests 79 4.5 EMERGENCY CORE COOLING SYSTEM AND REACTOR l
BUILDING COOLING SYSTEM PERIODIC TESTING 92 4.5.1 Emergency Core Cooling Systems 92 4. 5.' 2 Reactor Building Cooling Systems 95 4.6 AUXILIARY ELECTRICAL SYSTEM TESTS 100 4.7 REACTOR CONTROL ROD SYSTEM TESTS 102 4.7.1 Control Rod Drive System Functional Tests 102 3
4.7.2 control R M rogram Verification 104
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4.8 EMERGENCY FEEDWATER PUMP TESTING 105 4.9 REACTIVITY ANOMALIES 106 4.10 CONTROL ROOM EMERGENCY VENTILATION AND AIR CONDITIONING SYSTEM SURVEILLANCE 107 4.11 PENETRATION ROOM VENTILATION SYSTEM SURVEILLANCE 109
.4.12 HYDROGEN RECOMBINERS SURVEILLANCE 109b 4.13 EMERGENCY COOLING POND
.110a 4.14 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 110b 4.15 AUGMENTED INSERVICE INSPECTION' PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT 110c l
Amandment No. M,34,M,34,44, M,M, 11 M,44, M, M, MG, M-7,4M, M4 199
LIST OF FIGURES Numb 3 Titio Pogs 3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION COOLDOWN LIMITATIONS 20e 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H,0 33 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.8.1 SPENT PUEL POOL ARRANGDENT UNIT 1 59c 3.8.2 MAXIMUM BURNUP VS INITIAL ENRICHMENT FOR REGION 2 STORAGE 59d 3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110be 1
4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 4.18.3.a.3 110o2 5.4-1 ANO-1 FFSR LOADING PATTERN 116a f
Amendment No. M,M,H,4M,H4, iv M7, M9,444, H4, 199 I
3.6 REACT *R BUILDING Applic-bility
, Applies to the operability of the reactor building.
l Objective To assure reactor building operability.
l Specification 3.6.1 The reactor building shall be operable whenever all three (3) of the l
following conditions exists a.
Reactor coolant pressure is 300 psig or greater.
b.
Reactor coolant temperature is 200*F ce greater.
c.-
Nuclear fuel is in the core.
With the reactor building inoperable, restore the reactor building to operable status within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.6.2 Reactor building integrity shall be maintained when the reactor coolant system is open to the reactor building atmosphere and the requirements for a refueling shutdown are not met.
The provisions of Specification 3.0.3 are not applicable.
3.6.3 Positive reactivity insertions which would result in the reactor being suberitical by less than 1% Ak/k shall not be made by control rod motion or boron dilution whenever reactor building integrity is not in force. The provisions of Specification 3.0.3 are not applicable.
3.6.4 The reactor shall not be taken critical or remain critical if the reactor building internal pressure exceeds 3.0 psig or a vacuum of 5,5 inches Hg.
With the reactor critical, restore the containment pressure to within its limits within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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3.6.5 Prior to criticality following a refueling shutdown, a check shall be made to confirm that all nanual. reactor building isolation valves which should be closed are closed and locked, as required.
.The provisions of Specification 3.0.3 are not applicable.
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f Amendment No. 64, 199 54 1
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. If, while tha racetor is critical, a reactor building isolation-valve is determined to be inoperable in a position other than the I
closed position, the other reactor building isolation valve (except for check valves) in the line shall be tested to insure operability.
If the inoperable valve is not restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the operable valve will be closed.
Bases Included in reactor building operability are both the reactor building integrity as defined in Specification 1.7 and the reactor building structural integrity.
Structural integrity limitations as described in the ANO Containment Inspection Program ensure the reactor building will be maintained comparable to the original design standards.throughout the facility life span.
Visual and other required examinations of tendons, anchorages and surfaces are performed periodically in accordance with station procedures. These procedures embody applicable requirements of the 1992 Edition with the 1992 Addenda of Section XI, Subsection IWL of the ASME Boiler and Pressure vessel Code as set forth in j
Au crx 50.55a (g) (6) (ii) (B). Any degradations exceeding the Containment Inspection Program acceptance criteria during inspection surveillances will be reviewed under an engineering evaluation within 60 days.of the completion of the inspection to determine what impact the degradation has on overall containment operability, if any.
The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there will be no pressure buildup in the reactor building if the reactor coolant system ruptures.
The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.
The reactor building is designed for an internal pressure of 59 psig and an external pressure 3.0 psi greater than the internal pressure. The design external pressure of'3.0 psi corresponds to a margin of 0.5 psi above the differential pressure that could be developed if the building is sealed with an internal temperature of Il0*F and the building is subsequently cooled to an internal temperature of less than 50*F.
When reactor building integrity is established, the lindts of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur.
REFERENCE FSAR, Section 5.
l' Amendment No. 64, 199 55
"g Bases (1)
The reactor building is designed for an internal pressure of 59 psig and a I
steam-air udxture temperature of 285'F.
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The peak calculated reactor building pressure for the design basis loss of coolant accident, Pa, is 54 psig. The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.
The reactor building will be periodically leakage tested in accordance with the Reactor Building Leakage Rate Testing Program. These periodic testing requirements verify the reactor building leakage rate does not exceed the l
assumptions used in the safety analysis. At s 1.0 La the offsite dose i
consequences are bounded by the assumptions of the safety analysis.
During I
the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and s 0.75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L.
REFERENCE (1) FSAR, Sections 5 and 13.
l Amendment No. m,4 M,4 M,444 80 Next page is 92 l
199 I
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m-6.12.4 Rsector Building Inspsetion R3 port l
6 6.12.4.1 Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the ANO Containment Inspection Program shall undergo an engineering evaluation within 60' days of the completion of the inspection surveillance.
The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed.
The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions.to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations, i
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9 Amendment No. G, 43, 94, 444 146 199
6.12.5' Spacial Raports Special~ reports shall be. submitted to the Administrator of the appropriate Regional; Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of.the applicable reference specifiention.
a.
Deleted l
b.
Inoperable Containment Radiation Monitors, Specification 3.5.1, Table 3.5.1-1..
c.
' Dele".ed d.
Steam Generator Tubing Surveillance - Category C-3 Results,'
Specification 4.18.
.e.,
Miscellaneous Radioactive Materials' Source Leakage Tests,' Specification 3.12.2.
'f.
. Deleted g.
Deleted h.
Deleted 1.
Deleted
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Degraded' Auxiliary Electrical Systems, Specification 3.7.2.H.
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~ Inoperable Reactor Vessel Level Monitoring Systems, Table 3.5.1-l 1.
Inoperable Hot Leg Level Measurement Systems, Table 3.5.1-1 m.
Inoperable Main Steam Line Radiation Monitors, Specification 3.5.1, Table 3.5.1-1.
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, Amendment No. 88, M8,4M,444, Ma, 146a t
-us,448,.199 i