ML20091D257
| ML20091D257 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/31/1992 |
| From: | Larkins J Office of Nuclear Reactor Regulation |
| To: | Entergy Operations |
| Shared Package | |
| ML20091D259 | List: |
| References | |
| DPR-51-A-158, GL-86-010, GL-88-012, NPF-06-A-132 NUDOCS 9204090054 | |
| Download: ML20091D257 (49) | |
Text
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ENTERGY OPERATIONS, 1HC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO. 1 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 158 License No. DPR-51 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated October 15, 1991, as supplemented March 13, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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:
(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 Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
9204090054 920331 PDR ADOCK 05000313 P
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. 2.
Accordingly, the license is amended, as indicated in the attachment to this license amendment, and Paragraph 2.C.(8) of Facility Operating License No. DPR-51 is hereby amended to read as follows:
(8) FIRE PROTECTION E01 shall implement and maintain in effect all provisions of the approved Fire Protretion Program as described in Amendment 9A to the Safety Analysis Report and as approved in the Safety
]
Evaluation dated March 31, 1992, subject to the following provision:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
3.
In addition, the license is also amended by changes to the Technical
$pecifications as indicated in the attachment to this 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 Amendment No.
158, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
4 The license amendment is effective as of 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects 111, IV, and V Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: March 31, 1992
ATTACHMENT TO LICENSE AMENOMENT NO.158 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Revise the following pages of the License and 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.
REMOVE PAGES INSERT PAGES License page 5 License page 5 53d 53d 53e 53e 66m 66m 66n 66n 660 660 66p 66p 66q 66q 110p 110p 1104 110q 110r 110r 110s 110s 110t 110t 110u 110u 110v 110v 110w 110w 118 118 122 122 146b 146b N
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. (7) Secondary Water Chemistry Monitorino A secondary water chemistry monitoring program shall be implemented to minimize steam generator tube degradation. This program shall include:
1.
Identification of a sampling schedule for the critical parameters and control points for these parameters; 2.
Identification cf the procedures used to measure the values of the critical porameters; 3.
Identification of process sampling points; 4
Procedures for the recording and management of data; 5.
Procedures defining corrective actions for off-control point chemistry conditions; and 6.
A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate a corrective action.
(8)
FIRE PROTECTION E01 shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in Amendment 9A to the Safety Analysis Report and as approved in the Safety Evaluation dated March 31, 1992, subject to the following provision:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
3.
This license is effective as of the date of issuance and shall expire at midnight, May 20, 2014 FOR THE ATOMIC ENERGY COMMISSION Original Signed by:
A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing
Attachment:
Appendices A and B - Technical Specifications l
Date of Issuance: May 21, 1974 Amendment No.158
s e
3.5.5 E. ire Detection Instrumentation l
DELETED Amendment No. EP, f), 57, JJE 158 53d
Table 3.5-5 SAFETY-RELATED AREAS PROTECTED BY HEAT / SMOKE DETECTORS DELETED l
l l
Amendment No. p 158 53e
4 4
- E 3.17 Fire Suonression Water System l
DELETED l'
i-l Amendment No. 37, 37, J;g 158 66m
- 3.18-FIRE SUPPRESS 20N' SPRINKLER SYSTEM l
DELETED.
. Amendment No. Jf. 13, 57. 115 158 66n
a 3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS l
DELETED (u
Amendment No. Sf, 37, ;;p 158 66o
3, t.0
-TIRE HOSE STATIONS DELETLD Amendment No. 30,57 158 66p
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- '3.21
. TIRE BARRIERS
$ELETED l
Amendment No. $$, J7, 117 158 66q
- 4.19 FIRE DETECTION INSTRUMENTATION-DELETED.
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Amendment No. 39, 54 158 110p
4,20 Tir e Sur>nteision Wat e r Sy s t em DELETED
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I Amendment No. pp 158 110q
e DELETED Amendment No pp 158 110r
HAltE DELETED l
Amendment No. 3r 158 110s
4.21 SPRINKLER SYSTEMS DELETED l
l Amendment No. 35,J22, 176 158 110t
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4.22 Control Room and Auxiliary Control Room Halen Systems DELETED l.
i Amendment No. 39 158 110u
4.23
. Fire Hose Stations DELETED s
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Amendment No. JE 158 110v
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4.24 Fire Barriers l
DELETED Amendr.ent No. 39, 127 158 110w
Table 6 2-1 ARKANSAS NUCLEAR ONE ti1EUiUM SHIFT CREW COMPOSITION (!
UNIT 1 LICENSS COLD AND FEFUELING CATEGORY-ABOVE COLD SHUTDOWN SHUTDOWNS SOL 2
1*
OL 2
2 NON-LICENSED 3
None required
- Does not include the Ifcensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising refueling operations after the initial. fuel loading.
- Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to L restore the shif t crew composition to within the minimum requirements of
. Table 6,2-1.
Additional Requirements:
1.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
2.
At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown'and during recovery from reactor trips.
3.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
4.
All refueling operations after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
5.
DELETED Amendment No. If, )@, 25, 57 118 Jf6 158 3
h.
.Perforoancs of special rovices, invostigations, or analysca-and reports thereon as requested by the Plant Manager, ANO-1, General.ianager, Plant Operations or the Safety Review Committee.
i.
Review of the Plant Security Plan and implementing procedures and submittal of recommended changes to the General Manager, Plant Operations and the Safety Review Committee.
j.
Review of the Emergency Plan and' implementing procedures and submittal of recommended changes to the General Manager, Plant Operations and the Safety Review Committee.
k.
Review of changes to the Offsite Dose Calculation Manual and the Process Control Program.
1.
Review of changes to the Fire Protection Program and implementing procedures and submittal of recommended changes to the General Manager, Plant Operations and Safety Review Committee.
AUTHORITY 6.5.1.7.
The Plant Safety Committee shall:
a.
Recommend in writing their approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President, Operations ANO and the Safety Review Committee of disagreement between the PSC and the Plant Manager, ANO-1 or the General Manager, Plant Operations; however, the Geners' Manager, Plant Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The Plant Safety Committee shall maintain written minutes of each PSC meeting that, at a minimum, document the results of all PSC activities performed-under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Plcnt Manager, ANO-1, General Manager, Plar.t Operations -and Chairman of the Safety Review Committee.
6.5.2 Safety Review Committee (SRC)
TUNCTION 6.5.2.1 The Safety Review Committee shall function to provide independent review and audit of designated
- activities in the areas of:
a.
nuclear power plant operations b.
nuclear engineering c -.
chemistry and radiochemistry Amendment No. 16,34,37,57,57,22, 122 ft,19,it?,11F,151,1f7 158
4
- h. Inoperable Fire Detection Instrumentation
- 1. Inoperable Fire Suppression Systems J. Degraded Auxiliary Electrical Systems,' Specification 3.7.2.H.
- k. Inoperable Reactor Vessel Level Honitoring Systems. Table 3.5.1-1
- 1. Inoperable Not Leg Level Heasurement Systems. Table 3.5.1-1 Amendment No. 118. 131, 158 146b
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'b L k UNITED STATES i
NUCLEAR REGULATORY COMMISSION hj!fpf,/
- S WASHINGTON, o C. 20%6 gs, ENTERGY OPERATIONS. INC.
DOCKET NO. 50 368 ARKANSAS NUCLEAR ONE. UNIT NO. ?
Ar4ENDMENT TO FACILITY OPERATING LICENSC Arendment No.132 License No. NPT-6 1.
The Nuclear Regalatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee)datedOctober 15, 1991, as supplemented March 13, 1992, 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; B.
The facility will operate in conformity with the application, the ptovisior of the Act, and the rules and regulations of +he Comission:
C.
There is rearonable assurance:
(1) that the activities authorized by this amendment mn be conclucted without endangering the health and safety of the public, ano (ii) that such activities will be conducteo in compliance with the Comission's regulations; D.
The iss;, nce of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this asendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
3 l
. 2.
Accordingly, the license is amended, as indicated in the attachment to this license amendment, ead Paragraph 2.C.(3)(b) of Facility Operating License No NPF-6 is herevy amended to read as follows:
(b) Fire Protection-E01 shall implement and naintain in effect all provisions of the approved fire protection program as described in Amendment 9A to the Safety Analysis Report and as approved in the Safety Evaluation dated March 31, 1992 subject to the following provision:
The licsosee may n.ake changes to the approved fire protection program without prior approval of the Cournission only if those changes would not adversely affect;the ability t> achieve and maintain safe shutdown in the event of a fire.
3.
In addition, the license is also amended by changes to the Technical Specifications as-indicated in the attachment to this license anendment, and s'aragraph 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.132, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
o 4
The license amendment is effective as of 90-days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y
John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects 111, IV, and V Office of Nuclear Reactor Regulation r
Attachment:
Changes to the License and Technical Specifications Date of Issuance:
March 31, 1992 l
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i ATTACHMENT TO LICENSE AMENDMENT NO 11p FACILITY OPERATING LICENSE NO. NPT-6_
DOCKET NO. 50-368 Revise the following pages of the License and the Appendix "A" Technical Specificationt 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 License pages 4 and 5 License pages 4 and 5 3/4 3-43 3/4 3-43 3/4 3-44 3/4 3-44 3/4 7-29 3/4 7-29 3/4 7-30 3/4 7-30 3/4 7-31 3/4 7-31 3/4 7 32 3/4 7-32 3/4 7-33 3/4 7-33 3/4 7-34 3/4 7-34 3/4 7-35 3/4 7-35 3/4 7 36 3/4 7-36 3/4 7-37 3/4 7-37 B 3/4 3-3 B 3/4 3-3 B 3/4 7-6 B 3/4 7-6 B 3/4 7-7 B 3/4 7 7 6-2 6-2 6-5 6-5 6-7 6-7 6-19 6-19
4 (b) fire Protection E01 shall implement and maintain in effect all provisions of the approved fire protection program as described in Amendnent 9A to the Safety Analysis Report and as approved in the Safety Evaluation dated March 31, 1992
, subject to the following provision:
The licensee may make changes to the approved fire protection program without prior approval of the Connission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
Amendment No.132
-S-i (c) Less ihan Four Reactor Coolant Pump Operation E01 shall not operate the reactor in operational Modes 1 and 2 with fewer than four reactor coolant pumps in operation, except as allowed by Special Test Exception 3.10.3 of the facility Technical Specifications.
2.C.(3)(d)DeletedperAmendment 24,6/19/81.
(e) AP&L shall complete the following modifications by the indicated dates in accordance with the staff's findings as set forth in the fire protection evaluation report, NUREG-0223 " Fire Protection Safety Evaluation Report."
implementation Cates for Prooosed Modifications Applicable Section of NUREG-0223 Date 3.1 Portable Radio Communication Equipment March 31, 1979 3.2 Separation of Power Cables in Manholes 3.3 Protection from Water Spray 3.4 Protection of Redundant Cables in the MCC Room (2096-M)
December 30, 1978 3.5 Protection of Redundant Cables in the Hallway - Elevation 372(2109-U) 3.6 Protection of Redundant Cables in the CableSpreadingRoom(2098-L) 3.7 Protection of Redundant Cables in the SwitchgearRoom(2100-Z) 3.8 Protection of Redundant Cables in the Electrical Equipment Room (2091-BB)
September 30, 1978 Amendment No.132 l
l L
INSTRUMENTATION E1EI_ DETECTION INSTRUMENTAIIDS LIMITING CONDITION TOR OPEE/J1DN DELETED ARKANSAS - UNIT 2 3/4 3-43 Arnendinent No, 22132
PLANT SYSTEMS 3/4.7.10 r1RE st'PrREssioN sysIIts EIRE SlfP11ESS10N VATIJL.EJJIIB LlBill.ECOSDillD.Ulf DITJM198 DELETED t
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' ARKANSAS - UNIT 2 3/4 7-29 Amendment No. 132 i
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PLANT SY11I!iS EURVEILLANCE REQ 1!1REMENTLLCsALinggd)
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ARKANSAS - UNIT 2 3/4 7-31 Amendment No. 132
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g FLANT SYSTEMS SPRAY AND/OR SPRINKLIE_SYSTDiS LIMITING CONDITION FOR 01EK4 HQF DE1.ETED
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. ARKANSAS - UNIT 2 3/4 7-33 Amendment No. 132
PLANT SiSTEMS TIRE HOSE STATIONS MM1 TING CDSDLUDLIQJLQ[EE&llQN DELETED
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ARKANSAS - UNIT 2 3/4 7-35 Amendment No. 132
l INSTRU,':ENTATION
[] ' DETECTION INS.IfWlifNIAIJDS
). IING CONDIT1h M R,OPERAI10N l
DELETED ARKANSAS - UNIT 2 3/4 3-43
/.mendment No.22-132
TABLE 3.3-11 TfRE DETECTION INSTRUMENIS DELETED i
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l ARKANSAS - UNIT 2 3/4 3-44 Amendment No. 23. 17 132
' EL, ANT SYSTEMS 3/4.7.10 TIRE SUPPRESSION SYSTEMS TIRE SUPPRESElDN WATER SYSILM LIM 1U!iLC050lIIDS ICL0fEMHON DELETED L
i I
ARKANSAS - UNIT 2 3/4 7-29 Amendment No. 132 l
~
R NT SYSTEMS l
ACTION (Continued]
DELETED s.
l ARKANSAS - UNIT 2 3/4 7-30 Amendment No.112
PLANT SYSTEMS SURVEILLANpE REOUIF1.T.SI M G M Dued)
DELETED 9
ARKANSAS - UNIT 2 3/4 7-31 Amendment No. 132
PLANT SYSTEMS SURVE_1LL Qf M Epq11ERiENTS (continued)
DELETED l
d ARKANSAS - UNIT 2 3/4 7-32 Amendment No.132
PLANT SYSTEMS SPRAY AND/.01._EffdNELER SYSTEMS m ggJ pSDIT10N FOR OJ1 EAT 10N DELETED ARKANSAS - UNIT 2 3/4 7-33 Amendment No. 132
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SURVEILLANCE EEOUIREMENTS DELETED f.
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ARKANSAS - UNIT 2 3/4 7-34 Amendment.
No.132
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PLANT SYSIL!iS TIRE HDEL_EIAllfliS LIMITING CQ!iR111QGl@EftllD!{,,_
DELETED ARKANSAS - UNIT 2 3/4 7-35 Amendment No. 132
1 IABLt 3.7-7 T_lTJ._l[QSE STATIONS DELETED L
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~ ARKANSAS - UF.IT 2 3/4 7-36 Amendment No.132
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ILANT SYSTEliS 3/4.7.11 FIRE PARRIERS Illi1ILE C28.DHIEUCE_DELEAURS DELETED ARKANSAS - UNIT 2 3/4 7-37 Amendment No. 99,1:
_ PLANT SYSTEMS 3/4.7.12 SPENT FUEL POOL STRUCTURAL INT [GRITY MblIl!Kt10@lRQN FOR OPEf1110N 3.7.12 The t,tructural integrity of the spent fuel pool shall be maintsined in accordance with Specification 4.7.12.
APPLICABILITY:
Whenever irradiated fuel assemblies are in the spent fuel pool.
ACTION:
a.
With the structural integrity of the spent fuel pool not conforming to the above requirements, in lieu of any other report, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days of a determination of such non-conformity, b.
The provisions of Specification 3.0.2 are not applicable.
EfiXMMANS[22dPdMIS 4.7.12.1 Inspection Frequencies - The sttuctural integrity of the spent fuel pool shall be determined per the acceptance criteria of Specification 4.7.12.2 at the following frequencies:
a.
At least once per 92 days after the pool is filled with water.
If no abnormal degradation or other indications of structural distress are detected during five consecutive inspections, the inspection internal may be extended to at least unce per 5 years.
l b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following any seismic event which actuates or should have actuated the seismic monitoring instrumentation of Specification 3.3.3.3.
4.7.12.2 Acceptance Criteri_a - The structural integrity of the spent fuel pool shall be oetermined by a vitual inspection of at least the interior and exterior surfaces of the pool, the struts in the tilt pit, the surfaces of the separation walls, and the structural slabs adjoining the pool walls.
This visual inspection shall verify no changes in the concrete crat:k patterns, no abnormal degradation or other signs of structural distress (i.e, cracks, bulges, out of plumbness, leakage, discolcrations, efflorescence, etc.).
ARKANSAS - UNIT 2 3/4 7-38 Amendment No. 91,117
L, INSTRUMENTt.? ION PASES 4
3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameterr, to menitor and assess these variables following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97
" Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident." Dece.mber 1975 and NUREG-0578. "TMI-2 Lessons Learned Task Torce Status Report and Short Tena Recommendations."
The Reactor Vessel Level Monitor is provided as a means of indicating level in the reactor vessel during accident conditions.
A minimum of two operable level sensors in the upper plenum region and one operable level sensor in the dome region are required for RVLMS channel operability.
When Reactor Coolant Pumps are running, all except the dome sensors are interlocked to read " invalid" due to flow induced variables that may offset the sensor outputs.
If the equipment is inaccessible due to health and industrial safety concerns (for example, high radiation area low oxygen content of the containment atmosphere) or due to physical location i
of the fault (for example, probe failure in the reactor vessel), then operation may continue until the next scheduled refueling outage and a report flied.
3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release.
This capability is required to protect control room personnel and is consistent l
with the recommendations of Regulatory Guide 1.95.." Protection of Nuclear l
Power Plant Control Room Operators Against an Accidental Chlorine Release." February 1975.
l 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION 1
l DELETED l.l-ARKANSAS - UNIT 2 B 3/4 3-3 Amendment No. 22. 29 (9 223. 132
PLANT SYSTEMS PAsrs following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying t' control room to 5 ren or less whole body, or its equivalent. This limita is consistent with the requirements of General Design Criteria 19 o' pendix "A",
2/4.7.8 SHOCK SUPPRES.10RS ($NU1BERS)
All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained daring and following a seismic or other event initiating dynamic loads.
Snobbers excluded from this inspection prograu are those installed on nonsafety-related systems and then only if their faliure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
The visuoi inspection frequency is based upon maintaining a constant level of snubber protection to systems.
Therefore, the required inspection interval varies based upon the number of INOPERABh5 anubbers found during the previous inspection in proportion to the sizes of the various snubber populations or categories and the previous inspection interval as specified in NRC Generic Letter 90-09, " Alternative Re Inspection Intervals and Corrective Actions"quirements For Snubber Visual Inspections performed before that intarval has elapsed may be used as a r ew reference point to determine the next inspection.
How e ve tt, the result of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.
Any inspection whose results require a shorter inspection interval will override the previous schedule.
When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design featuras directly related to rejection of the snubber by visual inspection, or are simi*arly located or exposed to the same environmental conditions such as temperature, radiation and vibration.
When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of f ailure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation is performed to determine whether or not the enubber mode of failure has ieparted a significant effect or degradation on the supported component or system.
If a review and evaluation of an INOPERABLE snubber is performed and documented to justify continued operation and provided that all design criteria are met with the INOPERABLE snubber, then the INOPERABLE anubber would not need to be restored or replaced.
ARKANSAS - UNIT 2 B 3/4 7'5 Amendment No. AZ,129
PLANT SYSTEMS PASES To provide further assurance of snubber reliability, a representa-tive sample of the installed snubbers will be functionally tested during plant shatdowns at 18 month intervals.
These tests will include stroking of the snubbers to verify proper piston movement, lock-up and bleed.
Observed failures of these sample snubbers will require functional testing of addit')nal units. To minimize personnel exposures, snubbers installed in areas which have high radiation fields during shutdown or in especially difficult to remove locations may be exempted from these functional testing requirements provided the OPERABILITY of these snubbers was demonstrated during functional testing at either the completion of their fabrication or at a subsequent date.
3f4.7.9 SEALED SOURCE CONTAMINATION The limitetions on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
This limitation will ensure that leakage from byproduct, source, and speci&1 nuclear material sources will not exceed allowable intake values.
3/4.7.10 FIRE SUPPRESSION SYSTEMS DELETED ARKANSAS - UNIT 2 B 3/4 7-6 Amendment No.132
PLANT SYSTEMS BAsrs 3/4.7.11 FIRE BARRIERS DELETED 4
3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY The reinforcing steel in the walls of_the spent fuel pool was erroneously terminated into the front face instead of the rear face of the adjoining walls during construction of the spent fuel pool. Therefore, the specified structural integrity inspections of the spent fuel pool are required to be performed to ensure that the pool remains safe for use and that-it will adequately resist the imposed loadings.
If no abnormal degradation is observed during the first five inspections, the inspection interval for subsequent routine inspections may be extended to at least once per 18 months or longer if justified by observed performance of the
-pool.
l, 1>
ARKANSAS - UNIT 2 B 3/4 7-7 Amendment No. 99,132 1
I.
l
I c.
At least two licensed Operators shall be present in the control room during reactor start-up, scheduled t. actor shutdown and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor, All CORE ALTERATIONS shall be directly supervised by eit. er a e.
licensed Senior Reactor Operator or Senior Reactor Opera,or Limited to Fuel Handling who has no other concurrent responsibilities during this operetion.
f.
DELETED g.
Administrative control shall be established to limit the amount of overtime worked by plant staff performing safety-related functions.
These administrative controls shall be in accordance with the guidance provided by the NRC Policy Statement on working hours (Generic Letter No. 82-12).
h.
The Manager, Operations and Shift Supervisor shall hold a senior reactor operator license.
l 5
ARKANSAS - UNIT 2 6-2 Amendment No. 11,25,27,52, 11,ff,fl, 76 silf 132
AD31HIEIEATIVE CONTEDLS l
6.3 UNIT STATT 00ALITICATJQRS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the designated radiation protection manager, who shall meet or exceed +he qualifications of Regulatory Guide 1.8, Septembe'r 1975, and (2) the Shift Technical Advisor who shall have a bachelors degree, or equivalent in a scientif!c or engineering discipline with specif fr. training in plant design, and response and analysis of the plant for transsents and accidents, fJ_IRAltillW 6-A retraining and replacement training program for the unit staff shall bs
...aained under the direction of the Manager, Training and Emergency Planning and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CTR Part 55.
6.4.2 DELETED 6.5 FEVIEW AND A.UD11 Md PLANT SATETY C0tilTTEE (PSC)
TUNC1108 6.5.1.1 The Plant Safety Committee shall function to advise the General Manager, Plant Opr.ons and Plant Manager, AND-2 on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The Plant Safety Cc.mittee shall be composed of eight members of ANO onsite management organization (except as discussed under 6.5.1.3) at the superintendent level or above. The PSC Chairman shall ensure that adequate expertise is present during meetings to evalisate material before the PSC.
In addition, the General Manager, Plant Operations shall designate in writing a PSC Chairman and at least one Alternate Chairman.
6.5.1.3 If Core Protection Calculator (CPC) Software is being reviewed a nuclear software expert shall be present as a voting member.
If one of the members of the Plent Safety Committee meets the qualification requirements-for this position, the requirement to have this member is satisfied.
This membership may be filled by two appropriately qualified individuals who shall ballot with a single combined vote. Generic qualifications for this membership shall be as follows:
ARKANSAS - UNIT 2 6-5 Amendment No. ),J2,J/,
29,19,11,73,f 5, 9F,ilf,119 132
I A _s-il A0wn'!5TR ATIVE CONTD0Q i
Dit I W vidual The Nuclear Software Expert shall have as a minimum a Bachelor's degree in Science or Engineering, Nuclear preferred (in accordsn;e with ANSI N18.1).
In addition, he shall have a minimum of four years of technical experience, ef which a minimum of two years shall be in Nuclear Engintering and a n.irimun shall be in Software Engineering.
(Software Engimeering is that branch of science and technology which deals with the de'tign and use of software.
Software Engineering is a discipline directef to the production and modification of computer programs that are correct, efficient, flexible, enintainable, and understandable, in reasonaH e time spans, and at reasonable costs.) The two years of technical experience in Software Engineering r.ay be general software experience not necessarily related to the software of the Core Protectien Calculator System.
One of these two years of experience shall be with certified computer programs.
l Twe Individuals One of the individuals shall meet the requirements of the Nuclear Engineering portion of the above. The second individual shall have a Bar.helor of Scierre degree (cligital computer specirility) and meet the Software Engineering e quirements of the above.
The membership (the Nuclear Software Expert er the Digital Computer Specialist) shall be knowledgeable of the Core Protection Calculator System with regard to:
a.
The software modules, their interactions with eat;h other and with the data base, b.
The relationship between ;erator's redule inputs and the trip variables.
c.
The relationship between sensor input signals and the trip
- variable, d.
The design basis of the Core protection Calculator System.
e.
The a:p-oved software change procedure and documentation requin.nents of a sof tware change.
f.
The iccvHty of the computer memory and access procedures to the remory.
ARKANSAS-UNIT 2 64a Amendment No. 22,114
)
~ _ _ _. -
q e
CWSTJ4ID'E CONTR_OLS f.
Review of all REPORTABLE EVENTS.
g.
Review of facility operations to detect potential nuclear safety hazards.
h.
Performance of special reviews, investigations or analyses, and reports thereon as requested by the Plant Manager, ANO-2, General Manager, Plant Operations or the Safety Review Committee.
f.
Review of the Plant Security Plan a:.d implementing procedures and submittal of recommended changes to the General Manager, Plant Operatfors and the Safety Review Committec J.
Review of the Emergency Plan and implee.<rti,w rocedures and e
submittel of recommended changes to the ':c v..a1 Manager, Plant Operations and Safety Review %,.ittee.
k.
Review of charges to the Or fsi;c Dose Calculation Manual and Process Control Program.
1.
Review of changes to the Fire Protection Program and implementing, procedures and submittal of recommended changes to the General Manager, Plant Operations and Safety Revice Committee.
/,UTHORITY
- 6. 5.1. 8 -
The Plant Safety Committee shall:
a.
Recommend in writing their approval or disapproval of items considered under 6.5.1.6(a) through (d) above, b.
Render determinations in writing with regard to wheLher or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question, c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President,
. Operations ANO and the Safety Review Committee of disagreemen*
between the PSC and the Plant Manager, ANO-2 or the General h
- ger, Plant Operatione; however, the General Manager, Plant Operations shall have responsibility for resolution of such disagreements l
pursuant to 6.1.1 above.
RECORDS i
6.5.1.9 The Plant Safety Committee shall maintain written minutes of each PSC j
meeting that, at a minimum, document the results of all PSC activities
-performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Plant Manager, ANO-2, General Manager, Plant Operations and Chairman of the Safety Review Committee.
ARRANSAS - UNIT 2 6-7 Amendment No. 5,J/,25,52,fD, 77,rJ,91, 98,117.119 132
ASWIN!$TCATIVE CONTROLS f.5.2 LATETY REVIEV C2ww1TTEE (5Rti FUNET10N 6.5.2.1 The Safety Review Committee shall function to provide independent review and audit of designated activities in the areas of:
a.
nuclear power plant operations b.
nuclear engineering c.
chemistry and radiochemistry d.
metallurgy e,
instrumentation and control f.
radiological safety g.
mechanical and electrical engineering h.
quality assurance practices C0wr051 TION 6.5.2.2 The Safety Review Committee shall be composed of a Chairman and eight to twelve members which collectively have the experience and competence rsquired by AN51/ANS-3.1-1981 to review problems in the areas specified in Section 6.5.2.1, a h.
The Vice President, Operations ANO shall designate, in writing, the Chairman and all SRC members.
The Chairman shall designate, in writing, the alternate Chairman in the abstnce of the SRC Chairman.
ARKANSAS-UNIT 2 6-8 Amendment No. 5, 27, 21, 52, 7), 97, 78,114
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Adminis1rator of the Regior.a1 office within the time period specified for each report.
T.iese reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications
- a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Seirmic Monitoring Instrumentation, Specifica-tion 3.3.3.3.
]
- c. Inoperable Meteorological Monitoring Instrumentation. Specifica-tion 3.3.3.4.
- d. Seismic event analysis, Specification 4.3.3.3.2.
e.
Inoperable Fire Detection Instrumentation
- f. Inoperable Fire Suppression Systems
- g. Deleted.
- h. Padioactive Effluents, Specifications 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.2.5, and 3.11.3.
This report shall include the following:
1)
Description of occurrence.
z 2)
Identify the cause(s) for exceeding the limit (s) 3)
Explain corrective action (s) taken to mitigate occurrence.
4)
Define action (s) taken to prevent recurrence.
5)
Summary of consequence (s) of occurrence.
6)
Describe levels exceeding 40CFE190 in accordance with 10CFR20.405(c).
- 1. Inoperable Containment Radiation Monitors, Specification 3.3.3.1.
- j. Steam Generator Tubing Surveillance -- Category C-3 Results.
Specification 4.4.5.5.
- k. Maintenance of Spent Fuel Pool Structural Integrity, Specification 3.7.12.
ARKANSAS - UNIT 2 6-19 Amendment No. (0.(1,91, 92,132
m ADM7NISTRATIVE CONTROLE
- 1. Radiological Enrironmental Monitoring Sample Analysis, Specification 3.12.1.
- m. Unplanned Offsite Release during one hour period of 1) more than 1 curie of radioactive material in liquid affluents, 2) more than 150 curies of noble gas in gaseous affluents, or 3) more than 0.05 curies of radiciodine in gaseous affluents. This report shall be submitted within 30 days of the occurrenc9 of the event and shall include the following information:
1.
Description of the occurrence.
2.
Identify the cause(s) of exceeding the limit (s).
3.
Explain corrective action (s) taken to sitigate occurrence.
4.
Define action (s) taken to prevent recurrence.
5.
Summary of the consequence (s) of occurrence,
- n. Inoperable Reactor Vessel Level Monitoring System (RVLMS),
Specification 3.3.3.6, Table 3.3-10 Iten 14.
SEMI-ANNUAL RADI0 ACTIVE ErTLUENT RTT.T ASE REPORT
- 6.9.3 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operations shall be submitted within 60 days after January 1 and July 1 of each year.
l 1
- A single submittal say be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste system, the.
submittal shall specify the releases of radioactive material from each unit.
I l
ARKANSAS - UNIT 2 6-19s Amendment No, pl. 74,123
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