Text

Proposed Tech Specs for Licenses DPR-51 & NPF-6,relocating Noted Requirements to Reduce Costs for Licensees by Changing Requirements W/O Necessarily Requiring License Amend,Per GL-95-10
	ML20128L281
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/02/1996
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20128L278	List: 0CAN109602, Application for Amends to Licenses DPR-51 & NPF-6,reducing Costs for Licensees by Allowing Change to Requirements W/O Necessarily Requiring License Amend,Per GL 95-10; ML20128L281;
References
	GL-95-10, NUDOCS 9610110371
	Download: ML20128L281 (58)
	v • d • e

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1 PROPOSED TECHNICAL SPECIFICATION CHANGES i

9610110371 961002 PDR ADOCK 05000313 PDR P

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ANO-1

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'r 3.5.1.7 Tha Decay Hast Ramoval System isolation valva closure setpoints ohnll be equal to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in the auction line. The re, lief valve setting for the DHR system shall be equal to or less than 450 psig.

3.5.1.8 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 Feedwater shall be operable (as determined by Specification 4.1.a and item 35 of Table 4.1-1) at greater than 50 reactor power.

(May be bypassed up to 10% 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 5% 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 inoperable 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

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

If one channel of the NI/RPS is in naintenance bypass, only the corresponding channel of EFIC may be bypassed.

3.5.1.12 The Containment High Range Radiation Monitoring instrumentation, shall be operable with a minimum measurement range from 1 to 10 t

R/hr.

Amendment No, M,4,69,M,G4,404, 42a

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'r 3.5.1.13 D21stad l

3.5.1.14 TheMainSteamLineRadiationMonitoringInstrumentat{onshall 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 EFIC system which are bypassed at cold shutdown conditions shall have the following minimum operability conditions:

" low steam generator pressure" initiate shall be operable when a.

the main steam pressure exceeds 750 psig.

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

" main feedwater pumps tripped" initiate shall be operable when c.

neutron flux exceeds 10% power.

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

t 1

i Amendment No. 4M,MG,M3,4-74, 42b 1

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't Powar is normally supplisd to ths control rod driva mecheniems from two separate parallel 480 volt sources.

Redundant trip devices are employed 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 i

starting voltage protection as well as long term running voltage protection.

The 4.16 KV undervoltage relay setnoints are based on the allowable starting voltage plus maximum system volte r drops to the motor terminals, which allows i

approximately 78% of motor rated voltage at the motor tenninals. The 460V undervoltage relay setpoint is based on long term motor voltage requirements plus the maximum feeder voltage drop allowance resulting in a 92% 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 Following 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 thermocouples (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'se the approach to and o

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 dome 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 ndnimum 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 content 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.

Amendment No. M,69, M,444,446,444,-144 43b

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To support loss of main feedwater analyses, steem lins /feedwater lina break analyses, SBLOCA analyses, and HUREG-0737 requirements, the EFIC system is designed to automatically initiate EFW when:

1.

all four RC pumps are tripped 2.

both main feedwater pumps are tripped 3.

the level of either steam generator is low 4.

either steam generator pressure is low l

5.

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

The EFIC 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 logic:

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 EFIC 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 Amendment No, M 4,4,74, 43c

_ =. _ _ _

i Table 3.5.1-1 (cont'd)

OTHER SAFETY RELATED SYSTEMS

[

(Cont'd) 1 2

3 4

5 No. of Operator n<: tion channels Min.

Min.

if conditions of No. of for sys-operable degree of column 3 or 4 Functional Unit channels tem trip channels redundancy cannot be met 2.

Pressurizer level channels 2

N/A 2

1 Note 10 3.

Emergency Feedwater flow channels 2/S.G.

N/A 1

0 Note 10 4.

RCS subcooling margin monitors 2

N/A 1

0 Note 10 5.

Electromatic relief valve flow monitor 2

N/A 1

0 Note 11 6.

Electromatic relief block valve 1

N/A 1

0 Note 12 position indicator 7.

Pressurizer code safety valve flow 2/ valve N/A 1/ valve O

Note 10 monitors f

8.

Degraded Voltage Monitoring a.

4.16 KV Emergency Bus Undervoltage 2/ Bus 1/ Bus 2/ Bus 0

Note 14 b.

460 V Emergency Bus Undervoltage

l/ Bus 1/ Bus 1/ Bus 0

Notes 13, 14 9.

Deleted l

10. Containment High Range Radiation Monitoring 2

N/A 2

0 Note 20

11. Containment Pressure - High Range 2

N/A 2

0 Note 21

12. Containment Water Level - Wide Range 2

N/A 2

0 Note 21

Two undervoltage relays per bus are used with a coincident trip logic (2-out-of-2)

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i Amendment No. 60,60, 69,89, M,94, 4bd 4

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Table 3.5.1-1 (cont'd)

OTHER SAFETY RELATED SYSTEMS E

(cont'd) 1 2

3 4

5 No. of Operator action channels Min.

Min.

if conditions of No. of for sys-operable degree of column 3 or 4 Functional Unit channels tem trip channels redundancy cannot be met'

13. In core Thermocouples 6/ core quadrant N/A 2/ core quadrant 0

Note 22 (core-exit thermocouples)

14. Deleted

15. Reactor Vessel Level Monitoring System 2

N/A 2

0 Note 28, 29

16. Hot Leg Level Measurement System (HLLMS) 2 N/A 2

0 Note 28, 29

17. Main Steam Line Radiation Monitors 1/ steam line N/A 1/ steam line O

Note 30 Amendment No. 444,4-34,4M,443, 45d1

-m m.-

m

.-m :

Table 3.5.1-1 (cont'd) i 23.

With the number of operable Electronic (SCR) Trip relays one less than the total number of Electronic (SCR) Trip relays in a channel, restore the inoperable Electronic (SCR) Trip relay to operable status in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or place the SCRs associated with the inoperable Electronic (SCR) Trip relay in trip in the next hour. With two or more Electronic (SCR) Trip relays inoperable, place all Electronic (SCR) Trip relays associated with that channel in trip in the next hour.

This requirement does not apply to the Electronic Trip channels associated with Group 8 Regulating Power Supply.

r 24.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

I a.

Within 1 hour:

i 1.

Place the inoperable channel in the tripped condition, or 2.

Remove power supplied to the control rod trip device associated with the inoperable channel.

b.

One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing aad the

[

operable channel above may be bypassed for up to 30 minutes in any 24-hour period knen necessary' i

to test the trip breaker associated with the logic of the channel being tested. The inoperable channel above shall not be bypassed to test the logic of a channel of the trip system associated with the inoperable channel.

25.

With one of the Control Rod Drive Trip Breaker diverse trip features (undervoltage or shunt trip

~

attachment) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or place the breaker in trip in the next hour.

26.

Interrupts motor power to the Safety Groups of control rods only.

27.

Deleted t

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i Amendment No. 444,4-36,441, 45g l

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l Table 4.1-1 (Cont.)

t Channel Description Check Test Calibrate Remarks

29. High and Low Pressure NA NA R

Injection Systems:

Flow Channels

30. Decay heat removal S (l) (2)

M(1) (3)

R (1) Includes RCS Pressure Analog Channel system isolation valve (2) Includes CET Isolation Valve Position f

automatic closure and (3) At least once every refueling shutdown, interlock system with Reactor Coolant System Pressure greater than or equal to 200 psig, but less than 300 psig, verify automatic isolation of the decay heat removal system from the Reactor Coolant System on high Reactor Coolant System pressure.

31. Deleted

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32. Diesel generator M

Q NA protective relaying starting interlocks and circuitry f

33. Off-site power undervoltage W

R(1)

(1) Shall be tested during refueling i

and protective relaying shutdown to dee nstrate selective interlocks and circuitry load shedding interlocks function during manual or automatic trans-fer of Unit 1 auxiliary load to Startup Transformer No. 2.

i

34. Borated water storage W

NA R

tank level indicator l

35. Reactor trip upon loss M

PC R

of main feedwater circuitry l

t Amendment No. 4,4-9,60,60, G,94,442, 72 M,M, l

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Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks

36. Boric Acid Addition Tank I

a.

Level Channel NA NA R

b.

Temperature Channel M

NA R

37. Degraded Voltage Monitoring W

R R

38. Sodium Hydroxide Tank NA NA R

l Level Indicator l

t i

39. Incore Neutron Detectors M(1)

NA NA (1)

Check Functioning

40. Emergency Plant Radiation M(1)

NA R

(1)

Battery Check Instruments I

41. Reactor Trip Upon M

PC R

Turbine Trip Circuitry

42. Deleted I

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i Amendment No. 36,49,60,60,4+,91, 72a 444,4-36, t

s Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks I

43. ESAS Manual Trip Functions a.

Switches & Logic NA R

NA b.

Logic NA M

NA

44. Reactor Manual Trip NA P

NA

45. Reactor Building Sump Level NA NA R

46. EFW Flow Indication M

NA R

i Amendment No. M,-39,60, G, M, MO, 72b M, M4,

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Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks j

47. RCS Subcooling Margin D

NA R

Manitor

[

48. Electroma' tic Relief Valve D

NA R

Flow Monitor

49. Electromatic Relief Block D

NA R

Valve Position Indicator f

i 50._ Pressurizer Safety Valve D

NA R

Flow Monitor

[

.L

51. Pressurizer Water Level D

NA R

Indicator I

50. Deleted l

l

53. EEW Initiation i

i a.

Manual NA M

NA i

b.

SG Low Level, SGA or B S

M R

k c.

Low Pressure SGA or B S

M R

l d.

Loss of both MEW Pumps S

M R

[

]

and PWR > 10%

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^

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f I

Amendment No. 36,49,M,69,94,M6,4-74, 72b1 I

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMJTS_

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 DNBR MARGIN...................

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

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

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

3/4 2-9 3/4.2.8 PRESSURIZER PRESSURE......................

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 Pos t-Accicient Instrumentation..................

3/4 3-39 l

Fire Detection Ins trumentation.................

3/4 3-43 Radioactive Garcaus Efflusnt Monitoring I n s t r un en t a ti on...............,.............

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

3/4 3-54 i

ARKANSAS - UNIT 2 V

Amendment No. G4,49,43.4,443,

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INDEX i

l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

l SECTION PAGE l

j 3/4.4 REACTOR COOLANT SYSTEM l

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 l

3/4.4.5 STEAM GENERATORS........................................

3/4 4-6 l

l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAC"i l

Leakage Detection Systems...............................

3/4 4-13 Reactor Coolant System Leakage..........................

3/4 4-14 1

3/4.4.7 CHEMISTRY.................................

3/4 4-15 i

3/4.4.8 SPECIFIC ACTIVITY.......................................

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

3/4 4-22 i

Pressurizer.............................................

3/4 4 l l

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

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

3/4 5-1 l

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ARKANSAS - UNIT 2 VI Amentiment No. G9,60,6,

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i INDEX BASES

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SECTION PAGE l

3/4.0 APPLICABILITY..............................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL........................................

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS........................................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES..............................

, B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE........................................

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

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

B 3/4 2-2 3/4.2.4 DNBR MARGIN.....................>.......

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

B 3/4 2-4 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 PRESSURIZER PRESSURE....................................

B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 P ROTECTIVE 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 1

I ARKANSAS - UNIT 2 XI Amendment No. 34,34,60,

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THIS PAGE INTENTIONALLY LEET BLANK (Next page is 3/4 3-36) l l

ARKANSAS - UNIT 2 3/4 3-28 Amendment No. 63,M 4,.1M,g g,

. _ = -

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THIS PAGE INTENTIONALLY LEFT BLANK I

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ARKANSAS - UNIT 2 3/4 3-42 Amendment No, m,

=,

PLANT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup system satisfies the in place testing ac:eptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm 110%.

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 t10% during system operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by i

c.

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 Regalatory Position C.6.a of Regulatory Guide 1.52, l

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 i10%.

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 rem;ove 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%.

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

PLANT SYSTEMS 1

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 i

in accordance with Specification 4.7.12.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

With the structural integrity of the spent fuel pool not a.

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.

I SURVEILLANCE REQUIREMENTS 4.7.12.1 Inspection Frequencies - The structural integrity of the spent fuel pool shall be deteomined per the acceptance criteria of Specification 4.7.12.2 at the following frequencies:

i 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 l

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 should 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.

y This visual inspection shall verify no changes in the concrete crack patterns, no abnormal degradation or other' signs of structural distress (i. e, cracks, bulges, out of plumbness, leakage, discolorations, efflorescence, etc.).

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d ARKANSAS - UNIT 2 3/4 7-38 Amendment No. 91,444, I

i

3/4.3 XNSTRUMENTATION BASES 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 serve' 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 continuous 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 3/4.3.3.3 DELETED l

3/4.3.3.4 DELETED 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 General Design Criteria 19 of 10 CFR 50.

ARKANSAS - UNIT 2 B 3/4 3-2 Amendment No M,MO,Me,MG,

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XNSTRUMENTATION BASES 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 recommendations 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 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 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.

I I

ARKANSAS - UNIT 2 B 3/4 3-3 Amendment No. M, M,60,-1M, HG,

-.=- _ _ - -

e 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 initial calibration are used.

l ARKANSAS - UNIT 2 B 3/4 3-5 Amendment No. 60,

=.

- --- ~ - -.... _ _ - -

r 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 radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

MONTHLY OPERATING REPORT 6.9.1.6 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, l.

with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

l l

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 specification:

a.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b.

Deleted l

c.

Deleted l

d.

Deleted l

e.

Inoperable Fire Detection Instrumentation f.

Inoperable Fire Suppression Systems g.

Deleted

(

l l

ARKANSAS - UNIT 2 6-16 Amendment No. M,M,M,M,M G, 4M,

8 e

se f

l l

MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY) l l

l l

t l

l l

l

3.5.1.7 The Dscay Hast Rzmoval Systsm isolation valve closure setpoints shall bn equal to or less then 340 psig for one valva and equal to or less than 400 psig for the second valve in the suction line. The re, lief valve setting for the DHR system shall be equal to or less than 450 psig.

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

a.

The 4.16 KV emergency bus undervoltage relay setpoints 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 Feedwater 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 10% reactor power.)

2.

Reactor trip upon Turbine Trip shall be operable (as i

determined by Specification 4.1.a and item 41 of Table 4.1-1) at greater than 5% 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 inoperable 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 The cent::1 ::: ventil: tion chlorin: detection cy: tem instrument:ti:n ch:11 be cp; :ble and ::pabi of :tuating

nt:01 :::: isolation and filt: tion systems, with cle:r/tr+p

tpoint: :dju:ted t: ::tuct: at : chlorinc conec-t stien f 15pp:

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

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

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

Amendment No. 60,4,49,M,M,M4, 42a

e 3.5.1.13 gglggggThe :i:=ic ": nit: ring Instru=:nt:ti:n h:11 E: :p ::ble with :

minirux =c::urcsent ::ng: ; f 0. 01--

1.0 ; f : Tri ici Ti=:

i "ict :y *. ::1 : g::ph:, 0.05 1.0 ; f:: Tri: ici "::h

.. ::10 :g::ph;, :nd 0 2 5. ' '!: f:: Trisni:1 n :p:ns: Cpcstrum nc ::d ::.

j 3.5.1.14 The Main Steam Line Radiation Monitoring Instrumentation shall be operable with a minimum measurement range from 10-1 to 104 mR/hr, whenever the reactor is above the' cold shutdown condition.

3.5.1.15 Initiate functions of the EFIC 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.

" main feedwater pumps. tripped" initiate shall be operable when c.

neutron flux exceeds 10% power.

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

1

\\

Amendment No. -146,MO,M3, W, 42b i

a Powar is normally supplied to the control rod driva mzchenisum from two separate parallel 480 volt sources.

Redundant trip devices are employed 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 1

i 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 l

voltage plus maximum system voltage drops to the motor terminals, which allows approximately 78% of motor rated voltage at the motor terminals. The 460V undervoltate relay setpoint is based on long term motor voltage requirements plus the nu simum feeder voltage drop allowance resulting in a 92% 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 j

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

The OPEn?a!LITY cf th chierin: det:stien cycter cn ure: th:t cufficient eep bility 1: cvailable to p :;ptly dct t nd initist protective : tion i the event of :: :: ident:1 chierine ::1::::. Thi: ::p bility i: :: quired t: protect cent:01 :::: p ::enn:1 :nd i: consistent with th ::::;;:ndation: cf negulatory i

cuid: 1.05, "P: t :ti:n f Nucles: Peu : 01:nt Cent:01 nes: Operater: gninst :n

".::ident:1 Chlerine neles:," February 1975.

i The subcooled margin monitors (SMM), and core-exit thermocouples (CET), Reactor Vessel Level Monitoring System (RVLMS) and Hot Leg Level Measurement System 4

(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 diagnose the approach to and recovery from ICC.

Additionally, they aid in tracking reactor coolant I

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 reactor vessel during accident conditions. The channel operability of the RVLMS is defined as a ndnimum of three sensors in the upper plenum region and two sensors in the dome 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 content 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.

Amendment No. 60,69, G4,4M,4M,4M,444 43b

__.._._m.

Q The 0"En?."!LITY cf th: Ce$::1: M:nitorin; In:t :::ntatier en:ure: th:t euff6:icnt ::p:hility i: : :ilable t: p::;ptly det:: min: th: ::;nitud; cf a

i:=ic :: nt :nd ev:luct: th: :::p n:: cf th::: f: tur:: imp;; tent t: ::fety.

Thi: ::p hility i: ::quir d t permit ::xp:ri::n Of th: ::::ur:d :::p;n:: to th:t ::d i.- the d::ign b :i; for the f::ility t: determine if plant :hutdcun i:

quired pur:::nt t: App ndin

"?"

cf 100F" "::t 100.

The inst:: ent:ti:n i:

nzi t:nt with th: ::::n :nd: tion: ef S:fety Cuid: 12, "!r:trument:ti n f::

E::thquch,

,.......,...">.. npubli hed M :ch 10, 1971, 2nd 4t'nEC- 0 9 00 S :tien 3. ". 0,,

"S ci:=i c l

To support loss of main feedwater analyses, steam line/feedwater line break analyses, SBLOCA analyses, and NUREG-0737 requirements, the EFIC system is designed to automatically initiate EFW when:

l 1.

all four RC pumps are tripped 2.

both main feedwater pumps are 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 pressurr' The EFIC 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 logic:

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

i 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 EFIC 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 nanually, they will automatically reset.

Main feedwater pump trip bypass is automatically removed above 10% power.

REFERENCE FSAR, Section 7.1 FSAn, S : tic 2.'.S l

Amendment No. M4,4-7-7, 43c

s 1

Table 3.5.1-1 (cont'd)

OTHER SAFETY RELATED SYSTEMS s

(Cont'd) i 1

2 3

4 5

No. of Operator action channels Min.

Min.

if conditions of No. of for sys-operable degree of column 3 or 4 i

Functional Unit channels tem trip channels redundancy cannot be met i

2.

Pressurizer level channels 2

N/A 2

1 Note 10 i

3.

Emergency Feedwater flow channels 2/S.G.

N/A 1

0 Note 10 4.

RCS subcooling margin monitors 2

N/A 1

0 Note 10 i

5.

Electromatic relief valve flow monitor 2

N/A 1

0 Note 11 l

6.

Electromatic relief block valve 1

N/A 1

0 Note 12 f

position indicator l

7.

Pressurizer code safety valve flow 2/ valve N/A 1/ valve O

Note 10 monitors 8.

Degraded Voltage Monitoring i

a.

4.16 KV Emergency Bus Undervoltage 2/ Bus 1/ Bus 2/ Bus 0

Note 14 l

b.

460 V Emergency Bus Undervoltage

l/ Bus 1/ Bus 1/ Bus 0

Notes 13, 14 f

9.

Deleted Chlcrinc Oct ction Syst;ms 2

1 2

S M;te: 17, 19 l

6

10. Containment High Range l

Radiation Monitoring 2

N/A 2

0 Note 20 i

11. Containment Pressure - High Range 2

N/A 2

0 Note 21

12. Containment Water Level - Wide _ Range 2

N/A 2

0 Note 21 OTwo undervoltage relays per bus are used with a coincident trip logic (2-out-of-2) t i

f Amendment No. 60,60,69,89, M,94, 4Sd

.. -..-..=

c -.

s Table 3.5.1-1 (cont'd)

OTHER SAFETY RELATED SYSTEMS s

(Cont'd) 1 2

3 4

5 No. of Operator action channels Min.

Min.

if conditions of No. of for sys-operable degree of column 3 or 4 Functional Unit channels tem trip channels redundancy cannot be met

13. In core Thermocouples 6/ core quadrant N/A 1/ core quadrant 0

Note 22 (core-exit thermocouples)

14. DeletedScic=ic "cnitoring Inctrumentation c.

Trianicl Timc "ictory

".ccclcr:graphe u,i m n.

u.~__.-

m., m - o. n. n.,.,.....,

ContairT.cnt Sacc 'Sich, Elct. 335)

2.

.CS 9002, Unit 1 Tcp cf 1

" /.*

1 0

"st: 27 Centain=cnt, Slc.

531'5" r

b.

Tricxial "cch ".cccitrogrcph:

.e v. n os.,

,,__..:__..m__...

u,,s.,

n.

u..__.

~ ~......

r.,_.

~

.c i. u, _.~.

o.~s e. < =

n___

n._._..._._..

.c., u.... a

. rm n.

u..,__.m,,

m.

v. n.. o.~, n. o,......

C/S "ccctor Cavity, Elc; 355'3"

s..

. v.. n o s a. n, u _..:...,, _ _

m.c

u.,i.m

n..

u...__.-

,. ~ -

Ce n ta i r r,c n t, C1cv. 531'5"

.1 I

o With Unit I centrcl rccr indication /cr clar=

15. Reactor Vessel Level Monitorina System 2

N/A 2

0 Note 28, 29

16. Hot Lea Level Measurement System (HLLMS) 2 N/A 2

0 Note 28, 29

17. Main Steam Line Radiation Monitors 1/ steam line N/A 1/ steam line O

Note 30 P

r Amendment No. 4%,4M,4E M3.

45d1 l

[

1 f

L s

P. _ L.1 - -. t.,.

s___

eut Am. f. f. T9 M....h. P. PP. V.

h T"' T.h.P. T h -..f* P. N. D PV

. -..... m.

^____.__

.L...._.,__

u. d.

u..t...

.s

.- _. _J 2 a A _ _ _ _ r.

2

_____m,_

m_____

r..... _. :.....1.

t.r._. 4 6

.L...-...._.,_

a 2,-

L..-..._.,_

.-..J__..

L.._

.. ~

__J.

__a t

P.._.,__..,.

.~_e_.._.

5

.v.s...w m....,

e,cm, m.

.. -.... ~...

m___

, _m

,,c.

A,, e-em_

..._i....~_e1,

,e..~ _.,,____,. -..

T.~_.._,...

u.._._.__,__

m.

.. _ _ _. o.,

-,e

,fe

t. y.. v w.

Yu.3 av T._.._,.... u._w_ew_..___....

w y.a... -.

1r,, %

n a.

s....u~

e.._.__

E T.J. T.s a.w,t 8'l T ks e 8,

v

i. i v

.'l O

1. n 8

T_a.

vg r

g u..._.- _- ___

n.

,n

,,_m u.u sa wm.. v..

..v....vh.

(Numbers 15,16, and 17 have been moved to page 45d1 & this page will be deleted) l h

_ _ d_ _. 4 M.F _w.

,.wE,

,avw f,

......w.ii..

.y

A Table 3.5.1-1 (cont'd) 23.

With the number of operable Electronic (SCR) Trip relays one less than the total number of Electronic (SCR) Trip relays in a channel, restore the inoperable Electronic (SCR) Trip relay to operable status in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or place the SCRs associated with the inoperable Electronic (SCR) Trip relay in trip in the next hour.

With two or more Electronic (SCR) Trip relays inoperable, place all Electronic (SCR) Trip relays associated with that channel in trip in the next hour. This requirement does not apply to the Electronic Trip channels associated with Group B Regulating Power Supply.

24.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

Within 1 hour:

1.

Place the inoperable channel in the tripped cendition, or 2.

Remove power supplied to the control rod trip device associated with the inoperable channel.

b.

One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing and the operable channel above may be bypassed for up to 30 minutes in any 24-hour period when necessary to test the trip breaker associated with the logic of the channel being tested. The inoperable channel above shall not be bypassed to test the logic of a channel of the trip system associated 1

with the inoperable channel.

25.

With one of the Control Rod Drive Trip Breaker diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or place the breaker in trip in the next hour.

26.

Interrupts motor power to the Safety Groups of control rods only.

27.

DeletedWith cnc cr mort : iemic monitoring instrument in;pcr:ble for mar: than 30 day, pr;parc :nd ;;hnit :

Sp ici ncport to the Commiccion pursuant to Specificati n f.12.2 withir the n nt 10 day cutlining th: :::: cf th: malfunction and th: plans fcr :::tering the instrum:nt':: t OPERAELE statur.

The provision: cf Spccification 3.0.3 cr: not opplicable.

Amendment No. 444,4-34, M1, 45g

-..-~

~_--- _..- -.- -.

~.. - _ ~ ~. - ~ ~ -. -. -

i 4

Table 4.1-1 (Cont.)

i Channel Description Check Test Calibrate Remarks l

29. High and Low Pressure NA NA R

Injection Systems:

Flow Channels i

30. Decay heat removal S (1) (2)

M(1) (3)

R (1) Includes RCS Pressure Analog Channel system isolation valve (2) Includes CFT Isolation Valve Position I

automatic closure and (3) At least once every' refueling shutdown, f

interlock system with Reactor Coolant System Pressure j

grea%: than or equal to 200 psig, but less than 300 psig, verify automatic isolation of the decay heat removal system from the Reactor Coolant System on high Reactor Coolant System pressure.

I l

31. DeletedTurbin cvarsperdtrip (1) The previcien: ef Specification t.0. t i

=cchenis

not applicable.

r i

32. Diesel generator M

Q NA j

protective relaying starting interlocks and circuitry t

33. Off-site power undervoltage W

R(1)

(1) Shall be tested during refueling and protective relaying shutdown to demonstrate selective interlocks and circuitry load shedding interlocks function during manual or automatic trans-fer of Unit 1 auxiliary load to l

Startup Transformer No. 2.

34. Borated water storage W

NA R

[

tank level indicator

35. Reactor trip upon loss M

PC R

of main feedwater circuitry.

)

Amendment No. 4,M, M,60,67,94,M3, 72 M1, M3,

s Table 4.1-1 (Cont.)

t Channel Description Check Test Calibrate Remarks

36. Boric Acid Addition Tank i

a.

Level Channel NA NA R

b.

Temperature Channel M

NA R

37. Degraded Voltage Monitoring W

R R

j

38. Sodlum Hydroxide Tank NA NA R

Level Indicator L

39; Incore Neutron Detecte:s M(1)

NA NA (1)

Check runctioning

40. Emergency Plant Radiati a M(1)

NA R

(1)

Battery Check Instruments

41. Reactor Trip Upon M

PC R

Turbine Trip Circuitry

42. DeletedSci =ic " nitoring Instrument

Trinni:1 Time Mictory

'. ::1 r: graph i.

mee on.n.i,

n. _...m...

u.,,..

.e.m.

,, s, c_._. -_ -_ r_ m

~-.

n e_,__2_

si---

Contains:nt S ::

Elch, Elev. 335' (with Unit 1 control rc c indication' i

2.

..CS-S^02,

" nit 1

"(1)

S.'

Top of Contairment Elev. 531'5" Amendment No. M,M, M, M,4, M, 72a 4M,4-36,

t I

Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks u.

n. __ _ u..

".ccclc:cgraphe

1..

o v. n.

a o A. 9,,

e., _..:. n sr m se m n.

Centai-cnt Sacc Sich, i

e.s _..

,,c.cn

~.

I S

Tt_2..

.S sf a.

.S V. n. -

O S A. O.,

sf m n

n_..__.,.

.eu...

. 2 n,i e n.~_._

r.._.4...,

.e.e.s c3_..

k 3.

2X" S3'9, U.it 2 Tcp cf Contci-cnt, e.i _..

.s*

v w.

es,.c Triaxici "c p n c "pcetrum

";cc:dc:c

, v. n.

o..s e. n.,.....,

,,_m.

s, n.

n n

.e.i._ u,

......... ~...

n___

e,_..

,,e..c

, n,r e n__.

_a____

..... ~...,

43. ESAS Manual Trip Functions a.

Switches & Logic NA R

NA b.

Logic NA M

NA L

44. Reactor Manual Trip NA P

NA T

45. Reactor Building Sump Level NA NA R

46. EFW. Flow Indication M

NA R

e Amendment No. 24,49,60, 4,94,4-14, 72b 446,-140,

t i

Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks

47. RCS Subcooling Margin D

NA R

Monitor

48. Electromatic Relief Valve D

NA R

Flow Monitor

49. Electromatic Relief Block D

NA R

I Valve Position Indicator i

1

50. Pressurizer Safety Valve D

NA R

[

Flow Monitor

51. Pressurizer Water Level D

NA R

Indicator I

52. DeletedC:ntr:1 ",::- Chlorinc D:tccter 0 l

t

53. EFW Initiation I

a.

Manual NA M

NA 1

b.

SG Low Level, SGA or B S

M R

c.

Low Pressure SGA or B S

M R

(

d.

Loss of both MEW Pumps S

M R

and PWR > 10%

l F

b i

Amendment No. M,49, 54, 69,91, H&,4-M, 72b1

C p

/

E O

MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS (FOR INFO ONLY)

_m J

r INDEX 4

LIMITING CONDITIONS FOR OPERATION AND SURVEILIANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3,

.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 DNBR MARGIN....................................

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 1

3/4.2.7 AXIAL SHAPE INDEX..............................

3/4 2-9 3/4.2.8 PRESSURIZER PRESSURE...........................

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 Scismic In s t r ume n te.t i e n 3/' 3--34 Metes::1 gic 1 Instrumen W > mr~.

3/' 3-23 Remote Shutdown Instrumentation................

3 / 4."' - 3 6 Post-Accident Instrumentation..................

3/4 3-39 Chierin Detecti:n Systems.

3/' 3-'2 l

Fire Detection Instrumentation.................

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

3/4 3-43 Radioactive Liouid Effluent Monitorinc Instrumentation...........

3/4 3-54 ARKANSAS - UNIT 2 V

Amendment No. G4,60,4M,44,

e INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

'SECTION PAGE 3/t.3.3 (0:ntinuedF n:di:::ti;; Ligaid Effluent ":nitering

!n tru.T.cnt: tion (Moved to page V).

3/i 3 51 3/'.3.'

T"REINE OVEnSPEED rnOTECTION 3/' 3'59 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 S T EAM 3 EN E RATO RS........................................

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 4-25 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 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS..................................

3/4 5-1 ARKANSAS - UNIT 2 VI Amendment No. 2-9,60,6-3,

u a

i INDEX i

BASES SECTION PAGE 3/4.0 APPLICABILITY..............................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS I

3/4.1.1 BORATION CONTROL........................................

B 3/4 1-1 i

i

\\

j 3/4.1.2 BORATION SYSTEMS........................................

B 3/4 1-2 1

l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES..............................

B 3/4 1-3 1.

3/4.2 POWER DISTRIBUTION LIMITS l

3/4.2.1 LINEAR HEAT RATE........................................

B 3/4 2-1 j

3/4.2.2 RADIAL PEAKING EACTORS..................................

B 3/4 2-2 i

l 3/4.2.3 AZIMUTHAL POWER TILT....................................

B 3/4 2-2 j

3/4.2.4 DNBR MARGIN.............................................

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

B 3/4 2-4 1

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATU RE....................

B 3/4 2-4 5

]

3/4.2.7 AXI.'L SHAPE INDEX.......................................

B 3/4 2-4 4

3/4.2.8 P RES SU RI Z ER PRES S'.. ;E....................................

B 3/4 2-4 3/4.3 INSTRUMENTATION 1

3/4.3.1 PROTECTIVE INST RUMENTATION..............................

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 3/'.3.*

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l 1.

Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.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 fl0%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in i

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 i10% during system operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by c.

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.

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 j

i10%.

2.

Verifying that on a control room high radiation er high chlorir.: test signal, the 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 remove 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 i10%.

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

9 a.

PLANT SYSTEMS 3/4.7.12 SPENT FUEL POOL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 1

3.7.12 The structural integrity of the spent fuel pool shall be maintained 1

in accordance with Specification 4.7.12.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.

1 ACTION:

l With the structural integrity of the spent fuel pool not a.

I conforming to the above requirements, in lieu of any other report, i

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 l

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 j

inspection interval may be extended to at least once per 5 years.

j 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 should have actuated the seismic monitoring instrumentation-of Ep :i!! : tic 3.3.3.3.

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 patteras, no abnormal degradation or other signs of structural distress (i.e, cracks, bulges, out of plumbness, leakage, discolorations, i

efflorescence, etc.).

l l

ARKANSAS - UNIT 2 3/4 7-38 Amendment No. 91,444,

e 3/4.3 INSTRUMENTATION 1

BASES 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 perndt, whereas continuous 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 3/4.3.3.3 DELETEDEEISMIC INSTRUMENT

TION l

TS: OPER.*.EILITY cf th: ::ismic in:trumentation encurc; that cufficient ::pability i; cvailable to p mptly determine the magnitud cf

: ismi: cvent nd : alust: th: :::p:nce of th::: f::ture important t:

&&fety.

Thi: :pability i: :: quired t p rd t comp;;icen f the measu cd

p;ns to th:t u :d i.- th; design b;;i for the facility to determin; if pl:nt chutdown i; :: quired purcuant t:.'.pp;ndin ".'" of 10 CF" P :t 100.

The instrum ntation i: consistent with th: ::::mm:nd tion: Of S fety Cuide 12, " Instrumentation fc: E :thquakes," M :ch, 1071.

3/4.3.3.4 DELETEDMETEOnOLOCICAL INSTRUMENT

TION The OPER.'.EILITY cf the metes 1 gical inst :::nt-ation nsure that i

sufficient metecrelegicci det: i: : :11:b1: for :: tim: ting potential cadsstion de::: to the public :: : rc: ult of : cutin :: ::ident:1 cle:::

cf : dicactive met:ricl: to th:

t=::ph:::.

Thi: ::p:bility i: :: quired to cvalust: th; n :d f : initiating pretcetiv: measur;; to protect the h::lth :nd ::fety of the public csd i: ::nci; tent with th: ::::==:nds t-i-ene of negulatory cuide 1.22 "Oncit Metes::1 gie:1 0 :g::=:," Feburary 1972.

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 General Design Criteria 19 of 10 CFR 50.

ARKANSAS - UNIT 2 B 3/4 3-2 Amendment No. M,4M,HO,M3,

i e

INSTRUMENTATION BASES 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 recommendations 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 j

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 heal'..

and industrial safety concerns (for example, high radiation area, low oxygen content 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.

2/'.2.2.'

CMLon!ME EETECTION cy TEMS The OPEn.".E!L14Y cf th: chl:rin: detecti:n cycter encure: th t eufficient ::p:bility is sv:11:b1: t: p:: aptly det ct and initict protective ::ti n ir th: cvent f :n :::ident:1 chic in ::1 :::.

Thi:

ecpability i: ::quir:J to p :t :: cent :1 reem-per:cnnel and i: concietent with th: ::: xx:nd: tion; cf negulatory Cuid: 1.05, "P

t ction cf Mucles:

Peu : Plant Cent::1 n :: Op ::::::

'.g inct en '.::id:nt:1 Chlerin:

nel ::," Tehruary 1975.

2/'.2.2.S F!nE DETECTION IMcTnt' MENTATION OELETED ARKANSAS - UNIT 2 B 3/4 3-3 Amendment No. M, M,60,-1M,-1-34,

\\

1 5

l e<

. ~

INSTRUMENTATION BASES measurement assurance;acti.vities 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 initial calibration are used.

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ARKANSAS - UNIT 2 B 3/4 3-5 Arner.t.at No. 60,

O e,

,=

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 l

function of time for the duration of the specific activity above l

the steady-state levels and (5) The time duration when the l

specific activity of the primary coolant exceeded the radiciodine limit.

l HONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience j

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 specification:

a.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b.

Deleted!n p;::ble Sci:=i: ": nit:rin; :nctrumentati:n, Sp::ificatien 2.2 2.2.

l c.

Deleted 1nepc::b1: " t :::1:gie:1 ": nit:rin; Instrumentation, Sp;;ificati:n 2.2.2.t d.

DeletedSci:mi: :; nt 2n:1y ic, Sp::ific tic::

!.3.3.3.3.

l e.

Inoperable Fire Detection Instrumentation f.

Inoperable Fire Suppression Systems g.

Deletedv l

l ARKANSAS - UNIT 2 6-16 Amendment No. M,@,M,M, W,

W,

ML20128L281

Text

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