ML19340C039

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Proposed Tech Specs Re TMI-2 Lessons Learned Category a Items,Including Emergency Power Supply,Valve Position Indication,Instrumentation for Inadequate Core Cooling, Containment Isolation & Auxiliary Feedwater Sys
ML19340C039
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 10/31/1980
From:
ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19340C037 List:
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8011130205
Download: ML19340C039 (16)
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l (1) Faergency Power Supply T!equirements Af;0-1: Proposed T.S. 3.1.3.6 Atl0-2 : Proposed Changes to T.S. 3.4.4 (2) Valve Position Indication At'0- 1 : Proposed Changee o T.S. Table 3.5.1-1 Af;0-2: Proposed Changes to T.S. Table 3.3-10 (3) Instrumentation for Inadequate Core Cooling ANO-1: Proposed Changes to T.S. Table 3.5.1-1 A!!0-2 : Proposed Changes to T.S. Table 3.3-10 (4) Containment Isolation ANO-1: Existing T.S. Table 3.5.1-1 and T.S. 3.5.3 is adequate to meet the Afl0-1 design criteria Ah0-2: Existing T.S. 3.6.3 (5) Auxiliary Feedwater System ANO-1: Appropriate T.S. were submitted by AP&L letter dated February 12, 1980 ANO-2: Existing T.S. 3.7.1.2 and Table 3.3-3 (6) Shi f t Technical Advisor ANO-1: Proposed Changes to T.S. 6.3.1 and Table 6.2-1 j ANO-2: Proposed Changes to T.S. 6.3.1 and Table 6.2-1 "0111303,05

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3.1.3.6 The reactor shall not be made critical until at least 2 of the 3 cuergency powered pressurizer heater groups are operable.

With less than 2 of the 3 required heater groups operable, restore the required heater groups to operable st.at.us within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the required heater groups are not. restored to operable status wit.hin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in hot shutdown within t.he following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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If the r.hutdown margin required by Specification 3.5.2 is maint ained, there is no possibilit.y of an accidental criti-calit.y as a result of a decrease of coolant pressure.

The requirement for pressurizer bubble foriaation and specified water level when the reactor is less than one (1) percent subcritical will assure that the reactor coolan'. system cannot becorae solid in the event of a rod withdrawal accident or a start-up accident and that. the water level is above the mini-mum detectable level.

The requirement that 2 of the 3 emergency powered pressurizer heaters be operable orovides assurance that. sufficient heater capacity ( l 126 ku) is available to provide reactor coolant system pressure control during a loss of off-site power.

The requireuent that the safety rod groups be fully withdrawn bef ore criticality ensures shutdown capability during startup.

This does not prohibit rod latch confirmation, i.e., with-drawal by group to a maximuai of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.

The requirement for regulating rods being within their rod position limit.s ensures t hat the shut.down margin and ejected rod criteria at hot zero power are not. violated.

_REFEPENCES (1) FSAR, Section 3 (2) l'SAR, Secti on 3.2. 2.1.5 22

Table 3.5.1-1 (Cont.)

OTHER SAFETY PILATED SYSTEMS _

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

2. Steam line break instru-

> mentation control system (SLBIC).

(a) SLBIC Control & Logic 2 1 2 1 Notes 9, 5 Channels -

3. Pressurizer level channels 3 N/A 2 1 Note 10
4. Emergency Feedwater flow 2/S.G. N/A 1 0 Note 10 channels
5. RCS subcooling margin 2 N/A 1 0 Note 10 monitors
6. Electromatic relief valvu 2 N/A 1 0 Note 11 F flow monitor
7. Electromatic relief block 1 N/A 1 0 Note 12 valve position indicator
8. Pressurizer code safety 2/ valve N/A 1/ valve 0 Note 10 J valve flow monitors Notes: 1. Initiate a shutdown using normal operating instructions and place the reactor in the hot shutdown condition if the requirements of Columns 3 and 4 are not met within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2. When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

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3. When 1 of 2 intermediate range instrument channels is greater than 10 amps, hot shutdown is not required.
4. For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours, after which Note 1 applies.
5. If the requirements of Columns 3 or 4 cannot be met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, place the reactor in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6. The minimum number of operable channels may be reduced to 2, provided that the system is reduced to 1 out of 2 coincidence by tripping the remaining channel. Otherwise, Specification 3.3 shall apply.
7. These channels initiate control rod withdrawal inhibits not reactor trips at <10% rated power. Above 10% rated power, those inhibits are bypassed.
8. If any one compenent of a digital subsystem is inoperable, the entire digital subsystem is considered inoperable. Ilence, the associated safety features are inoperable and Specification 3.3 applies. .
9. The minimum number of operable channels may be reduced to one and the minimum degree of redundancy to zero for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after which Note 1 applies.
10. With the number of operable channels less than required, either restore the inoperable channel to operable status within 30 days, or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
11. With the number of operable channels less than required, isolate the electromatic relief valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise Note 9 applies.
12. With the number of operable channels less than required, either return the indicator to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or verify the block valve closed and power removed within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the block valve cannot be verified closed within the additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, de-energize the electromatic relief valve power supply within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l Table 4.1-1 (Cont.)

Channel Description Check Test Calibrate Remarks

47. RCS subcooling margin D N/A R monitor
48. Electromatic relief D N/A R valve flow monitor
49. Electromatic relief D N/A R block valee position indicator
50. I'ressurizer safety D N/A R valve flow monitor
51. Pressurizer water D N/A R level indicator

Table 4.1-2

- Minimum Equipment Test. Frequency Item Test Frequency

1. Control Rods Rod Drop Times of all Each Refueling Shutdown Full Length Rods 1/
2. Cont rol Rod Movement of Each Rod Every Two Weeks Above Cold Movement Shutdown Conditions
3. Pressurizer Code Setpoint One Valve Every 18 !!onths Safety Valves
4. Main Steam Safety Setpoint Four Valves Every 18 Months Valves
5. Refueling System Functioning Start of Each Refueling Interlocks Shutdown
6. React.or Coolant Evaluate Daily System Leakage
7. Emergency powered Power availability Daily Pressurizer IIcaters IIcater capacity Every 18 Months functional test
8. Reactor Building Functioning Every 18 Months Isolation Trip
9. Service Water Functioning Every 18 Months Systems
10. Spent. Fuel Functioning Every 18 Months Cooling System uhen irradiated fuel is in the pool.
11. Decay IIcat Removal Functioning Every 18 Months System Isolat. ion Valve Automatic Closure and Isolation System If Same as tests listed in Section 4.7

6.0 ADMINISTRATIVE CONT!iOLS 6.1 RESPONSIBILITY 6.1.1 The General Manager shall be responsible for overall facility ,

operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION 2

OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2.-l .

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figures 6.2-2A, B, C, and D. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1. .

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable position, except for (1) the Health Physics Supervisor who shail meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) tht. Shift Technical Advisor uho shall have a bachelor's degree or equivalcat in a scienti fic or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING l

6.4.1 A retraining and replacement training program for the facility staff shall be maintained and shall meet. or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appen-dix "A" of 10 CFR Part 55.

6.4.2 A training program for fire protection training shall be maintained and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975 with the exception of frequency of training which shall be six times per year.

6.5 REVIEW AND AUDIT 6.5.1 Plant Safety Committee (PSC) Function 6.5.1.1 The Plant Safety Committee shall function to advise the General Manager on all matters related to nuclear safety.

COMPOSITION l 6.5.1.2 The Plant Safety Committee shall be composed of the:

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Table 6.2-1 ARKANSAS NUCLEAR ONE M nit 1UM SHIFT CREW C0r! POSITION #

UNIT 1 COLD AND REFUELING LICENSE ADOVE COLD SIIUTDOWN SHUTDOWNS CATEGORY 1

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SOL 1

OL 2 1

HON-LICENSED 2

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1 None required l S111FT TECHNICAL AINISOR

  • Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising refueling operations after the initial fuel loading.
  1. Shift rew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to rest. ore the shif t crew composit. ion to within the minimum requirements of Table 6.2-1.

i Additional Requirements:

1. At. least one licensed Operator shall be in the cont.rol roorr. when l fuel is in the reactor.
2. At. least t_wo licensed Operators shall be present in the control l

room during reactor start-up, scheduled reactor shutdown and during I recovery f rorr. react or trips.

3. An individual qualified ia radiation protection procedures shall be on ait.e when fuel is in t.he reactor.
4. All refueling operations after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

M S. At least 5 individuals with fire protection training shall be maintained onsite at all times. These individuals shall not include the minimum shift crew necessary for safe shutdown of the unit (2 members) or any personnel required for ot.her essential functions during a fire emergency.

    • This change from 3 to 5 individuals will be implemented within 90 days of date of issuance of this license amendment.

REACTOR COOLANT SYSTElf PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer.shall be OPERABLE with a water volume of

< 910 cubic feet (equivalent to < 82% of wide range indicated 1cvel), and both pressurizer proportional heater groups shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

l_iCTION :

(a) With the pressurizer inoperable due to water volume > 910 cabic feet be in at least -l!OT SHUll)0WN with the reactor trip breakerr. open within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(b) With the prest.urizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the in-operable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 110T SilUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQU1REMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.4.2 The pressurizer proportional heater groups shall be determined to be OPERABLE:

(a) At least once per twelve hours by verifying emergency power is available to the heater groups, and (O At 1 cast once per 18 months by verifying that the summed power consumption of the two proportional heater groups is > 150 KW.

Table 3.3-10 POST-ACCIDENT 110NITORING INSTRUMENTATION MINIllut!

CIIANNELS INSTRUMENT OPERABLE

1. Containment Pressure 2
2. Containment Radiation Monitors'c 2
3. Pressurizer Pressure 2
4. Pressuri7er Water Level 2
5. Steam Generator Pressure 2/stea n generator
6. Steau Generator Water Level 2/ steam generat.or
7. Refueling Water Tank Water Leve] 2
8. Containment. Sump Water Level 2
9. Emergency Feedwater Flow Rate 1/ steam generator
10. Reactor Coolant System Subcooling 1 Margin Monitor
11. Pressurizer Safety Valve Flow !!onitor 1/ valve sfhis requirement may be satisfied by the use of portable radiation monitors equivalent in number to the minimum channels required OPERABLE until r.ucli time as the Category B portions of Item 2.1.8.B of NUREG 0578 must be implemented for ANO-2.

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Table 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREM CHANNEL CllANNEL CllECK CALIBRATION INSTRUMENT M R

1. Containment Pressure M R
2. Containment Radiation Monitors
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3. Pressurizer Pressure M R
4. Pressurizer Water Level M R
5. Steam Generator Pressure M R
6. Steara Generator Water Level M R
7. Refueling Water Tank Water Level M R
8. Containment Sump Water Level M R
9. Emergency Feedwater Flow Rate M R
10. Reactor Coolant System Subcooling Margin Monitor M R
11. Pressurizer Safety Valve Flow Monitor 1

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  • This requirement may be satisfied by the use of portable radiation monitors, and by substituting a source check for the channel check and by substituting an instrument calibration for the channel calibration until such time as the Category B portions of Item 2.1.8.B of NUREG 0578 must be implemented for ANO-2.

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Table 6.2-1 tlINIMUt! SilIFT CREW COMPOSITION //

LICENSE CATEGORY APPLICABLE MODES 1, 2, 3 & 4 5&6 SOL 1 li' OL 2 1 Non-Licensed Auxiliary Operator 2 1 Shift Technical Advisor 1 None Required

  • Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to
  • Fuel llandling, supervising CORE OPERATIONS. -

//Shif t crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty chift crew members provided immediate action is taken to restore the shift. crew composition to within the minimum requirements o f Table 6. 2-1.

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ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF (U!ALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Mealth Physics Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September, 1975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline witA specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under t.he direction of the General Manager and shall meet or exceed the r.equirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFit Part 55.

6.4.2 A training prop, ram for the Fire Brigade shall be maintained under the direct _ ion of the General Manager and shall meet or exceed the requirenents of Section 27 of the NFPA Code - 1975, except for Fire Brigade training sessions which chall be held at least quarterly.

6.5 REVIEW AN1) AUDIT l

6.5.1 PLANT SAFETY COMMITTEE (PSC)

FUNCTION 6.5.1.1 The Plant Safety Comaittee shall iunction to advise the General Manager on all matters related to nuclear safety.

COMPOSIT_ ION 6.5.1.2 the Plant Safety Committee shall be composed of the-Chairman: Operations and Maintenance Manager Member: Operations Superintendent Member: Technical Analysis Superintendent Member: Maintenance Superintendent Member: Instrumentation & Controls Superintendent Membe r: Plant Analysis Superintendent Member: Mealth Physics Supervisor Member: Nuclear Sof tware Expert

  • The General Manager shall appoint in writing an acting chairman in the absence of the Operations and Maintenance Manager.
  • See page 6-Sa

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INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrument ation ensures that suf-ficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recormnendations of Regulatory Guide 1.97, "Instru-mentation for Light-Water-Cooled Nuclear Plants to Assess Plant. Condi-tions During and Following an Accident," December 1975, and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report. and Short-Term Recom-mendations."

3/4.3.3.7 CllLORINE DETECTION SYSTE!!S The OPERABILITY of' the chlorine detection system ensures that. sufficient capability is available to promptly detect and initiate protective act. ion in the event of an accidental chlorine release. This capability is required to prot.ect control room personnel and is consistent with t he recour.cudaticas of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release,"

February 1975.

3/4.3.3.8 FIRE DETECT]ON INSTRUMENTATION i OPERABILITY of the fire detection instrumentation ensures that adequate l warning capability is available for the prompt detection of fires. This capability is required in order to detect. and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event. that a portion of the fire detection instrumentat. ion is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is rest.ored to OPERABILITY.

3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Pro-t ection f rom turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.

l REACTOR COOLANT SYSTEM 1

BASES Demonstration of the safety valves' lif t set. tings will occur only during shut.down and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydrau-lically solid system and is capable of accommodating pressure surges during operat ion. The steam bubble also protects the pressurizer code safety valves against. water relief. The steam bubble functions to relieve RCS pressure during all design transients.

The requirement that 150 kw of pressurizar her,ters and their associated cont rols be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of of fsite power condit. ion to maintain natural circulation at. HOT STAND-BY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensures that the structural inte;,rity of this portion of the RCS will be maintained. The program or inservice inspection of steam gen-erator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maint.ain surveillance of the conditions of the tubes in the event th L there is evidence of mechanical damage or progressive de-gradation due t o design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result. in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between t.he primary coolant system and the secondary coolant system (primary-to-secondary leakage = 0.5 GPM per steam generator). Cracks have a primary-to-secondary leakage less than this limit during operat. ion will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary to-secondary leakage of 0.5 CI'M per steam generator can readily be detected by radia-tion monitors of steam generator bloudown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

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