Text

Amends 192 & 191 to Licenses DPR-51 & NPF-6,respectively, Revising ANO-1 & 2 TSs by Relocating TS Requirements Related to Instrumentation from TS to UFSAR
	ML20236R384
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/13/1998
From:	William Reckley; Office of Nuclear Reactor Regulation
To:	; Entergy Operations
Shared Package
ML20236R386	List: ML20236R384; ML20236R387;
References
	DPR-51-A-192, NPF-06-A-191 NUDOCS 9807210422
	Download: ML20236R384 (33)
	v • d • e

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t UNITED STATES j

j NUCLEAR REGULATORY COMMISSION

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t WASHINGTON, D.C. 30606-0001

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ENTERGY OPERATIONS. INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.192 License No. DPR-51 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated October 2,1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (tne Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confom1ity with the application, the p,ovisions 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 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 iri accordance with 10 CFP Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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9907210422 980713 PDR ADOCK 05000310 P

PDR l

N 2-2.

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. DPR-51 is hereby amended to read as follows:

(2)

Technical Specifications l

The Technical Specifications contained in Appendix A, as revised through i

Amendment No.192, 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 at the facility within 30 days (including issuance of Technical Requirements Manual for use by licensee personnel). In addition, the licensee shallinclude the relocated information in the Updated Final Safety Analysis Report submitted to the NRC, pursuant to 10 CFR 50.71(e), as was described in the licensee's application dated October 2, 1996, and evaluated in the staffs safety evaluation dated July 13,1998.

FOR THE NUCLEAR REGULATORY COMMISSION William Reckley, Project M ger Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 13,1998

ATTACHMENT TO LICENSE AMENDMENT NO.192 FACILITY OPERATING LICENSE NO. DPR-5_1 DOCKET NO. 50-313 Revise the following pages of the Appendix "A" Technical Spe.ifications with the attached pages.

l The revise 1 pages are identified by Amendment number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 42a 42a 42b 42b 43b 43b 43c 43c 45d 45d 45d1 45d1 45d2 45g 45g 72 72 72a 72a 72b 72b 72b1 72b1

3.5.1.7 Th3 Dec:y H t Ramoval System locleticn vsiva cleauro cctp31nts chall b3 cqu:1 to or Icss then 340 psig for en3 v31v3 cnd cqual to or less than 400 psig for the second valve in the suction line. The reAief valve setting for the DHR system shall be equal to or less than 450 psig.

3.5.1.0 The degraded voltage monitoring relay settings shall be as follows:

The 4.16 KV emergency bus undervoltage relay setpoints a.

shall be >3115 VAC but <3177 VAC.

b.

The 460 V emergency bus undervoltage relay setpoints shall be >423 VAC but <431 VAC with a time delay setpoint of 8 seconds il second.

3.5.1.9 The following Reactor Trip circuitry shall be operable as indicated:

1.

Reactor trip upon loss of Main roadwater shall be operable (as determined by Specification 4.1.a and item 35 of Table 4.1-1) at greater than 5% reactor power.

(May be bypassed up to 106 reactor power.)

2.

Reactor trip upon Turbine Trip shall be operable (as determined by Specification 4.1.a and item 41 of Table 4.1-1) at greater than 56 reactor power.

(May be bypassed up to 45% reactor power.)

3.

If the requirements of Specifications 3.5.1.9.1 or 3.5.1.9.2 cannot be met, restore the jnoperable trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or bring the plant to a hot shutdown condition.

3.5.1.10 Deleted l

l 3.5.1.11 For on-line testing of the Emergency reedwater Initiation and Control (ETIC) system channels during power operation only one channel shall be locked into " maintenance bypass" at any one tima.

If one channel of the HI/RPS is in maintenance bypass, only the corresponding channel of Er1C may be bypassed.

3.5.1.12 The Containment High Range Radiation Monitoring instrumentation, shall be operable with a minimum umasurement range from 1 to 10 R/hr.

l l

l Amendment No. 40,4,4,M,H,M4,192 42"

____ _______ - _ A

o 3.5.1.13 Deleted l

3.5.1.14 TheMainSteamLineRadiationMonitoringInstrument3tgench311 be operable with a minimum measurement range from 10-to 104 mR/hr, whenever the' reactor is above the cold shutdown condition.

3.5.1.15 Initiate functions of the EPIC system which are bypassed at cold shutdown conditions shall have the following minimum operability conditions:

a. " low steam generator pressure" initiate shall be operable when the main steam pressure exceeds 750 psig,

b. " loss of 4 RC pumps" initiate shall be' operable when neutron flux exceeds 10% power.

c. " main feedwater pumps tripped" initiate shall be operable when neutron flux exceeds lot powe.t..

3.5.1.16 The automatic steam generator isolation system within EFIC shall be operable when main steam, pressure is greater than 750 psig.

Amendment No. 4M,MG,Ma,4w,192 42b 4

o

(

Power 10 normally supplied to the control red drive machenicas from two ocparato pSrc1131 480 valt ocurcos. Redund:nt trip d2vicas cro ampicy:d in each of these sources. If any one of these trip devices fails in the untripped state, on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested.

Four hours is ample time to test the remaining trip devices and, in many cases, make on-line. repairs.

The Degraded Voltage Monitoring relay settings are based on the short term starting voltage protection as well as long term running voltage protection.

The 4.16 KV undervoltage relay setpoints are based on the allowable starting voltage plus maximum system voltage drops to the motor tenminals, which allows approximately.7sa of motor rated voltage at the motor terminals. The 460V j

undervoltage relay setpoint is based on long term motor voltage requirements plus the maximum feeder voltage drop allowance resulting in a 926 setting of motor rated voltage.

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess.these variables during and following an accident. This capability is consistent with the recommendation of Regulatory Guide 1.97, " Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant Conditions *During and Folloking an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

l The subcooled margin monitors (SMM), and core-exit thermocouple (CET), Reactor vessel Level Monitoring System (RVLMS) and Hot Leg Level Measurement System (HLLMS) are a result of the Inadequate Core Cooling (ICC) instrumentation required by Item II.F.2 NUREG-0737; The function of the ICC instrumentation is to increase the ability of the plant operators to diagn'ose the approach to and recovery from ICC.

Additionally, they aid in tracking reactor coolant inventory. These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37 and are not required by'the accident analysis, nor to bring the plant to cold shutdown conditions. The Reactor vessel Level Monitor is provided as a means of indicating level in the reac' tor vessel during accident conditions. The channel operability of the RVLMS is defined as a minimum of three sensors in the upper plenum region and two sensors in the done region operable. 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. The channel operability of the HLLMS is defined as a minimum of one wide range and any two of the narrow range transmitters in the same channel operable.

If the equipment is inaccessible due to health and industrial safety concerns (for example, high radiation area, low oxygen cor. tent of the containment atmosphere) or due to physical location of the fault (for example, probe failure in the reactor vessel), then operation may continue until the next scheduled refueling outage and a report filed.

43b Amendment No. g 69,84,444,446,4M,444,

o To cuppsrt less of main feedwatcr entlysos, etcam lino /foedwater lino brock analyses, SBLOCA analyses, and NUREG-0737 requirements, the EFIC system is designed to automatically initiate EFW when:

1.

all four RC pumps are tripped 2.

.both main feedwater pumps ara tripped 3.

the level of either steam generator is low 4.,

either steam generator pressure is low 5.

ESAS ECCS actuation (high RB pressure or low RCS pressure)

The ErIC system is also designed to isolate the affected steam generator on a steam line/feedwater line break and supply EFW to the intact generator according to the following logics

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If both SG's are above 600 psig, supply EFW to both SG's.

If one SG is below 600 psig, supply EFW to the other SG.

If both SG's are below 600 psig, but the pressure difference between the two SG's exceeds 100 psig, supply EFW only to the SG with the higher pressure.

If both SG's are below 600 psig and the pressure difference is less than 100 psig, supply EFW to both SG's.

At cold shutdown conditions all EPIC initiate and isolate functions are bypassed except low steam generator level initiate.

The bypassed functions will be automatically reset at the values or plant conditions identified in Specification 3.5.1.15.

" Loss of 4 RC pumps" initiate and " low steam generator pressure" initiate are the only shutdown bypasses to be manually initiated during cooldown. If reset is not done manually, they will automatically reset. Main feedwater pump trip bypass is automatically removed above 10% power.

REFERENCE FSAR, Section 7.1 l

l l

1 l

Amendment No. 464,447, 192 43e l

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1 UNITED STATES j

NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30ssMicot 44. ;...,o ENTERGY OPERATIONS. INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.191 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 October 2,1996, 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: (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.

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.191, are hereby incorporated in the license. The licensee shall l

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 at the facility within 30 days (including issuance of Technical Requirements Manual for use by licensee personnel). In addition, the licensee shallinclude the relocated information in the Updated Final Safety Analysis Report submitted to the NRC, pursuant to 10 CFR 50.71(e),

as was described in the licensee's application dated October 2,1996, and evaluated in the staff's safety evaluation dated July 13,1998.

FOR THE NUCLEAR REGULATORY COMMISSION

)

1

}]b Wilham Reckley, Project r

Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: July 13,1998

ATTACHMENT TO LICENSE AMENDMENT NO.191 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain verticallines indicating the area of change. 'The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES V

V VI VI XI XI

Xil Xll 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-30 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 3/4 3-42 3/4 3-42

3/4 3-57 3/4 3-57 3/4 3-58

3/4 7-17 3/4 7-17 3/4 7-18 3/4 7-18

3/4 7-37 3/4 7-37 3/4 7-38 3/4 7-38 B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 B 3/4 3-5 B 3/4 3-5 f

6-15 6-15 i

6-16 6 '.6 l

l l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE...............................

3/4 2-1 3/4.2.2 RADIAL PEAKING FACTORS.........................

3/4 2-2 3/4.2.3 AZIMUTHAL POWER TILT...........................

3/4 2-3 3/4.2.4 DN R xRRGIN..................................

3/4 2-5 3/4.2.5 RCS FLOW RATE..................................

3/4 2-7 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE...........

3/4 2-8 3/4.2.7 AXIAL SHAPE INDEX..............................

3/4 2-9 3/4.2.8 PRES SURIZ ER PRES SURE...........................

3/4 2-10 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION.............

3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.............................

3/4 3-10 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...........

3/4 3-24 i

Remote Shutdown Instrumentation................

3/4 3-36 Post-Accident Instrumentation..................

3/4 3-39 Fire Detection Instrumentation.................

3/4 3-43 Radioactive Gaseous Effluent Monitoring Instrumentation.............................

3/4 3-45 Radioactive Liquid Effluent Monitoring Instrumentation..............................

3/4 3-54 ARKANSAS - UNIT 2 V

Amendment No. G4,40,u4,Ma,191

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE I

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOIANT LOOPS AND COOLANT CIRCULATION...........

3/4 4-1 3/4.4.2-SAFETY VALVES - SNUTDOWN................................

3/4 4-3 3/4.4.3 SAFETY VALVES - OPERATING...............................

3/4 4-4 3/4 4-5.

3/4.4.4 P RES SURI Z ER...................,..........................

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 ACTIVITY.......................................

3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System..................................

3/4 4-22 Pressurizer.............................................

3/4 4 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 i

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS..................................

3/4 5-1 ARKANSAS - UNIT 2 VI Amendment No. 49,M,44,191 a

____.___..________________________________,__________________j

o INDEX BASES PAGE SECTION B 3/4 0-1 3/4.0 APPLICABILITY..............................................

3/4.1 REACTIVITY CONTROL SYSTEMS B 3/4 1-1 3/4.1.1 BORATION CONTROL........................................

B 3/4 2-2 3/4.1.2 BORATION SYSTEMS........................................

3/4.1.3 MOVABLE CONTROL ASS EMBLI ES............................... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS B 3/4 2-1 3/4.2.3 LINEAR HEAT RATE........................................

3/4.2..

RADIAL PEAKING FACTORS..................................

B 3/4 2-2 3/4.2.3 AZIMUTHAL POWER TILT....................................

S'3/4 2-2 B 3/4 2-3 3/4.2.4 DNBR MARGIN.....................s.......................

B 3/4 2-4 3/4.2.5 RCS FLOW RATE...........................................

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE....................

B 3/4 2-4 3/4.2.7 AXIAL SHAPE INDEX.......................................

B 3/4 2-4 3/4.2.8 PRES S U RI Z ER PRES SURE....................................

B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION................

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..............................

B 3/4 3-2 I

I l

\\

ARXANSAS - UNIT 2 XI Amendment No. 44,M,H,191

THIS PAGE INTENTIONALLY LEFT BIANK (Next page is 3/4 3-36) l ARKANSAS - UNIT 2 3/4 3-28 Amenchment No. H,H4,4H,MS,191 -

me en o e

-*+~.n<sm

THIS PAGE INTENTIONALLY LEFT BLANK l

ARKANSAS - UNIT 2 3/4 3-42 Amendment No. M4,191

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PLANT SYSTEMS 4

3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING AND AIR FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 1 Two independent control room emergency air conditioning and air filtration systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2. 3 and 4.

ACTION:

With one control room emergency air conditioning or air filtration system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.6.1.1 Each control room emergency air conditioning system shall be

(

demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGERED TEST BASIS by:

1.

Starting each unit from the control room, and 2.

Verifying that each unit operates for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and maintains the control room air temperature < 84'F D.B.

b.

At least once pe: 18 months by verifying a system flow rate of 9900 cfm i 105.

4.7.6.1.2 Each control room emergency air filtration system shall be demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verffying that the system operates for at least 15 minutes.

b.

At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by:

ARKANSAS - UNIT 2 3/4 7 17

---.9

.2---------------------------------------------------------

e PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm 1106.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide '1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory, Guide 1.52, Revision 2, March 1978.

3.

Verifying a system flow rate of 2000 cfm 110% during system operation when tested in accordance with ANSI N510-1975.

c.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, M rch 1978, meets the laboratory testing a

criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the system at a flow rate of 2000 cfm ilot.

2.

Verifying that on a control room high radiation test signal, the l

system automatically isolates the control room within 10 seconds and switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

e.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks restove 299% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 2000 cfm 110%.

I i

I I

ARKANSAS - UNIT 2 3/4 7-18 Amendment No.191

4 PLANT SYSTEMS 3/4.7.11 FIRE BARRfERE LTMYTTNG CONDYTTON TOR OPTRATION DELETED 1(

t L

ARKANSA8 - UNIT 2 3/4 7-37 Amendment No. 99r132

PIANT SYSTEMS 3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.7.12 The structural integrity of the spent fuel pool shall be maintained l

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 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12.1 Inspection Frequencies - The structural 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 interval may be extended to at least once per 5 years.

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following any seismic event which actuates or s%.ald have actuated the seismic monitoring instrumentation.

l 4.7.12.2 Acceptance Criteria - The structural integrity of the spent fuel pool shall be determined by a visual 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 crack patterns, no abrormal degradatiori or other' signs of structural distress (i.e, cracks, bulges, out of plumbness, leakage, discolorations, efflorescence,'etc.).

ARKANSAS - UNIT 2 3/4 7-38 Amendment No. M,444,191

3/4.3 ZN97RLVENTATZON BA9ES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION

" The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

The PURGE as defined in the definitions section is a release under a purge permit, whereas continuoet ventilation is defined as operation of the purge system after the requirements of the purge permit have been' satisfied.

When securing the containment purge system to meet the ACTION requirements of this specification, at least one supply valve and one exhaust valve is to be closed, and the supply and exhaust fans secured.

3/4.3.3.2 DELETED l

3/4.3.3.3 DELETED l

3/4.3.3.4 DrIETED l

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with Generel Design Criteria 19 of 10 CFR 50.

ARKANSAS - UNIT 2

.B 3/4 3-2 Amendment No. M,M4,Ho,wa)91

INSTRUMENTATION BA3ES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the reconsoendations of Regulatory Guide 1.97,

" Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, *TMI-2 Lessons Learned Task Force Status Report and Short Term Reconsnandations. "

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 i

offset the sensor outputs. If the equipment is inaccessible due to health

)

and industrial safety concerns (for example, high radiation area, low l

oxygen content of the containment atmosphere) or due to physical location l

of the fault (for example, probe failure ih the reactor vessel), then operation may continue until the next scheduled refueling outage and a report filed.

I I

ARKANSAS - UNIT 2 B 3/4 3-3 Amendmer.t No. M,M,40,+H,4M,191 I

INSTRUMENTATION BASES measurement assurance activities with NBS.

These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the init.ial calibration are used.

l ARKANSAS - UNIT 2 B 3/4 3-5 Amendment No. 48,191.

L.

i 4(

$k ADMINT $TeATf E CONTROt$

ANNUALREPORy' 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March I of each year.

The initial report shall be submitted prior to March 1-of the year following initial criticality.

6.9.1.5.

Reports required on an annual basis shall include:

A tabulation on an annual basis for the number of station a.

and other personnel (including contractors) receiving expo,suresutility greater than 100 aren/yr and their a accordingtoworkandjobfunctions,pociatedmanremexposure e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling.

The dose assignment to various duty functiou estimates based on pocket dosimeter, TLD, or film badg4 r o be measurements.

individual total dose need not be accounted for.Small exposure In the aggregate, at least 80% of the total whole body dose received form external sources shall be assigned to specific major work functions.

b.

The complete results of steam generator tube inservice inspections performed during the report period (reference Specification

(

4.4.5.5.b).

Documentation of all challenges to the pressurf2er safety valves.

c.

d.

A diesel generator data report which provides the number of valid tests and the number of valid failures for each diesel generator.

The results of specific activity analysis in which the primary e.

coolant exceeded the limits of Specification 3.4.8.

information shall be included:

The following 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was(1) Reactor p exceeded; (2) Results of the last isotopic analysis for ridiciodine performed prior to exceeding the limit, results'of annlysis while licit was exceeded the results of one analysis after the radiciodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radiciodir,e concentrations; (3) Clean up system flow history y

A single subaittal may be made for a multiple odt station.

The submittal should combine those se'etions that av common to all units at the station.

1/

This tabulation supplomants the requirements of f20.407 of 10 CFR i

part 20.

l l

l ARKANSA$ - UNIT 2 6-15 AMEN 0 MENT NO. 5, #I,H,92

ADMINISTRATIVE CONTROLS starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state levels and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

MONTHLY b?"EnTING REPORT 6.J.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, office of Resource, Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

SPECIAL REPORTS 6.9.2.Special reports shall be submitted to the Administrator of the 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 specifications a.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b.

Delsted l

l c.

Deleted l

d.

Deleted l

e.

Inoperable Fire Detection Instrumentation f.

Inoperable Fire Suppression Systems g.

Deleted l

l l

ARKANSAS - UNIT 2 6-16 Amendment No. M,60,M,43,HG, M4,191 j

_ _ _.

ML20236R384

Text

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