Text

Proposed Tech Specs Implementing Selected Improvements as Described in GL 93-05
	ML18017A927
Person / Time
Site:	Harris
Issue date:	10/21/1999
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML18017A926	List: ML18017A925; ML18017A927;
References
	GL-93-05, NUDOCS 9910280146
	Download: ML18017A927 (193)
	v • d • e

ENCLOSURE 5 TO SERIAL: HNP-99-149 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT PROPOSED IMPLEMENTATIONOF NRC,GENERIC LETTER 93-05 TECHNICALSPECIFICATION PAGES ES-1 j 99i0280i46 99X02i PDR ADQCK 05000400 P PDR

REACTIVITY CONTROL SY MS LIMITING CONDITION FOR OPERATION ACTION Continued remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3. 1. 1. 1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; c) A power distribution map is obtained from the movable incore .

detectors and F~(Z) and F," are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) The THERMAL POWER level is reduced to less than or equal to 75K of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85K of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS

4. 1.3. 1. 1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. 1.3. 1.2 Each rod not fully inserted in the core shall be determined to be OPE LE by movement of at least 10 steps in any one direction at least once per 31 ays.

A,d,d.

SHEARON HARRIS - UNIT 1 3/4 1-15

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE RE UIREMENTS DIGITAL CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST IS RE UIRED

1. Containment Radioactivity--

a. Containment Ventilation R Q N 1, 2, 3, 4, 6 Isolation Signal Area Monitors

b. Airborne Gaseous Radioactivity

1) RCS Leakage Detection R 1, 2, 3, 4

2) Pre-entry Purge R

c. Airborne Particulate Radioactivity Qk bc'iW

1) RCS Leakage Detection w 1, 2, 3, 4

2) Pre-entry Purge R 9

2. Spent Fuel Pool Area--

Fuel Handling Building Emergency Exhaust Actuation

a. Fuel Handling Building Operating Floor--South Network

b. Fuel Handling Building Operating Floor--North Network SHEARON HARRIS - UNIT 1 3/4 3-54 Qmcnd+~$ No.

TABLE 4.3-3 Continued RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE RE UIREMENTS DIGITAL CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST ~IS RE UtREO

3. Control Room Outside Air Intakes

a. Normal Outside Air Intake Isolation S R Al 1

b. Emergency Outside Air Intake S R Al 1 Isolation--South Intake Q

c. Emergency Outside Air Intake S All Isolation--North Intake TABLE NOTATIONS With irradiated fuel in the Northend Spent Fuel Pool or transfer of irradiated fuel from or to a spent fuel shipping cask.

With irradiated fuel in the Southend Spent Fuel Pool or New Fuel Pool.

Whenever pre-entry purge system is to be used.

gg Prior to operation of pre-entry purge unless performed within the last 31 days.

SHEARON HARRIS - UNIT 1 3/4 3-55

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1227 cubic feet, equivalent to 92K of indicated span, and at least two groups of pressurizer heaters each having a capacity of at least 125 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within'6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3. 1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by e ing the heaters and measuring circuit power (kW) at least once per 2 days. Qc.4+

SHEARON HARRIS - UNIT 1 3/4 4-10 Amnlge~k h)O.

REACTOR COOLANT SYST OPERATIONAI LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2. 1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or Particulate Radioactivity Monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment sump inventory and Flow Monitoring System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 + 20 psig at least once per 31 days with the modulating valve fully open.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least er 18 months.

se.V

b. Prior t SHUTDOWN fo e

2 E 2 whenever the plant has been in COLD ours or more and if leakage testing has not been performed in ious 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

SHEARON HARRIS - UNIT 1 3/4 4-24

CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4. 1 Two independent containment hydrogen monitors shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.

b. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 .

SURVEILLANCE REQUIREMENTS 4.6.4. 1 Each hydrogen monitor shall be demonstrated OPERABLE by th performance of an ANALOG CHANNEL OPERATIONAL TEST at least once pe ~31 days.

and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:

a. Two volume percent hydrogen, balance nitrogen. and

b. Six volume percent hydrogen, balance nitrogen.

a~i~+

SHEARON HARRIS - UNIT 1 3/4 6-30 Amendment No. 47

CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS LIMiTING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE.

APPLICABILITY: MODES 1 and Z.

ACTION:

With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.

SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recomb r System shall be demonstrated OPERABLE:

a. At least once per w ~,~

6 months by verifying, during a Hydrogen Recombiner System unctional test, that the minimum heater sheath temperature increases to greater than or equal to 700 F within 90 minutes. Upon reaching 700'F. increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 kW, and At least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits,

2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and Verifying the integrity of all heater electrical ci rcuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

SHEARON HARRIS - UNIT 1 3/4 6-31 g~~ .I. <a.

PLA T SYSTEMS AUXILIARY FEEDWATER SYSTEM LIM/TING CONDITION FOR OPERATION 3.7. 1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor -driven auxiliary feedwater pumps, each capable of being powered from separate emergency buses. and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable. be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary p~( +

feedwater pump to OPERABLE status as soon as possible. (NOTE:

LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Following restoration of one AFW train, all applicable LCOs apply based on the time the LCOs initially occurred.)

SURVEILLANCE REQUIREMENTS 4.7. 1.2. 1 Each auxiliary feedw pump shall be demonstrated OPERABLE:

DcAM

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

Demonstrating that each motor-driven pump satisfies performance requirements by either:

a) Verifying each pump develops a differential pressure that (when temperature - compensated to 70 F) is greater than or equal to 1514 psid at a recirculation flow of greater than or equal to 50 gpm (25 KPPH), or b) Verifying each pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1259 psid at a flow rate of greater than or equal to 430 gpm (215 KPPH).

Ce.(~

SHEARON HARRIS - UNIT 1 3/4 7-4 Amendment No. 51

PLANT SYSTENS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREHENTS (Continued)

2. Demonstrating that the steam turbine - driven pump satisfies performance requirements by either:

NOTE: The provisions of Specification 4.0.4 are not applicable for entry into NODE 3.

a) Verifying the pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1167 psid at a recirculation flow of greater than or equal to 81 gpm (40.5 KPPH) when the secondary steam supply pressure is greater than 210 psig. or b) Verifying the pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1400 psid at a flow rate of greater than or equal to 430 gpm (215 KPPH) when the p4 secondary steam supply pressure is greater than 280 si 4o 4 Itdtbk op~

3. eri ying y ow position check that each valve (manual,

>>> power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its

~

correct position; and g+v

~

4. Verifying that the isolation valves in the suction line from the CST are locked open.

b. At least once per 18 months by:

1. Verifying that each motor-driven auxiliary feedwater pump starts automatically, as designed, upon receipt of a test signal and that the respective pressure control valve for each motor-driven pump and each flow control valve with an auto-open feature respond as required;

2. Verifying that the turbine-driven auxiliary feedwater pump starts automatically, as designed. upon receipt of a test signal. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; and Verifying that the motor-operated auxiliary feedwater isolation valves and flow control valves close as required upon receipt of an appropriate test signal for steamline differential pressure high coincident with main steam isolation.

>eih SHEARON HARRIS - UNIT 1 3/4 7-5 Amendment No. 87

REACTIVITY CONTROL SY MS LIMITING CONDITION FOR OPERATION ACTION Continued remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3. 1. 1. 1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; c) A power distribution map is obtained from the movable incore detectors and F~(Z) and F," are verified to be within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75K of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85K of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS

4. 1.3.1. 1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. 1.3. 1.2 Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days.

SHEARON HARRIS - UNIT 1 3/4 1-15 Amendment No.

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE RE UIREMENTS DIGITAL CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST IS RE UIRED

1. Containment Radioactivity--

i

a. Containment Ventilation S 1, 2, 3, 4, 6 Isolation Signal Area Monitors

b. Airborne Gaseous Radioactivity

1) RCS Leakage Detection S 1, 2. 3, 4

2) Pre-entry Purge S

c. Airborne Particul ate Radioactivity

1) RCS Leakage Detection S 1, 2, 3, 4

2) Pre-entry Purge S

2. Spent Fuel Pool Area--

Fuel Handling Building Emergency Exhaust Actuation

a. Fuel Handling Building S Oper ating Floor--South Network

b. Fuel Handling Building S Oper ating Floor--North Network SHEARON HARRIS - UNIT 1 3/4 3-54 Amendment No.

TABLE 4.3-3 Continued RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE RE UIREMENTS DIGITAL CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST ~IE RE UIIIEU

3. Control Room Outside Air Intakes

a. Normal Outside Air Intake Isolation S All

b. Emergency Outside Air Intake S All Isolation--South Intake

c. Emergency Outside Air Intake S All Isolation--North Intake TABLE NOTATIONS With irradiated fuel in the Northend Spent Fuel Pool or transfer of irradiated fuel from or to a spent fuel shipping cask.

With irradiated fuel in the Southend Spent Fuel Pool or New Fuel Pool.

Whenever pre-entry purge system is to be used.

g Prior to operation of pre-entry purge unless performed within the last 92 days.

SHEARON HARRIS - UNIT 1 3/4 3-55 Amendment No.

REA TOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1227 cubic feet, equivalent to 92K of indicated span, and at least two groups of pressurizer heaters each having a capacity of at least 125 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit power (kW) at least once per 18 months.

SHEARON HARRIS - UNIT 1 3/4 4-10 Amendment No.

REA TOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS I

4.4.6.2. 1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment Airborne Gaseous or Particulate Radioactivity Monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment sump inventory and Flow Monitoring System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 + 20 psig at least once per 31 days with the modulating valve fully open.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

SHEARON HARRIS - UNIT 1 3/4 4-24 Amendment No.

CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4. 1 Two independent containment hydrogen monitors shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 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 .

b. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 .

SURVEILLANCE REQUIREMENTS 4.6.4. 1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL .OPERATIONAL TEST at least once per 92 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:

a. Two volume percent hydrogen, balance nitrogen, and

b. Six volume percent hydrogen, balance nitrogen.

SHEARON HARRIS - UNIT 1 3/4 6-30 Amendment No.

CONTAINMENT SYSTEMS f

ELECTRIC HYDROGEN RECOMBINERS LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.

SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:

At least once per 18 months by verifying,'uring a Hydrogen Recombiner System functional test, that the minimum heater sheath temperature increases to greater than or equal to 700'F within 90 minutes. Upon reaching 700'F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 kW, and

b. At least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits,

2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

SHEARON HARRIS - UNIT 1 3/4'-31 Amendment No.

PLA T SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING."CONDITION FOR OPERATION 3.7. 1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency buses, and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. (NOTE:

LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Following restoration of one AFW train, all applicable LCOs apply based on the time the LCOs initially occurred.)

SURVEILLANCE REQUIREMENTS 4.7. 1.2. 1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

At least once per 92 days on a STAGGERED TEST BASIS by:

1. Demonstrating that each motor-driven pump satisfies performance requirements by either:

a) Verifying each pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1514 psid at a recirculation flow of greater than or equal to 50 gpm (25 KPPH), or b) Verifying each pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1259 psid at a flow rate of greater than or equal to 430 gpm (215 KPPH).

SHEARON HARRIS - UNIT 1 3/4 7-4 Amendment No.

PLANT SYSTEMS l'UXILIARYFEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Demonstrating that the steam turbine - driven pump satisfies performance requirements by either:

NOTE: The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

a) Verifying the pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than 'or equal to 1167 psid at a recirculation flow of greater than or equal to 81 gpm (40.5 KPPH) when the secondary steam supply pressure is greater than 210 psig, or b) Verifying the pump develops a differential pressure that (when temperature - compensated to 70'F) is greater than or equal to 1400 psid at a flow rate of greater than or equal to 430 gpm (215 KPPH) when the secondary steam supply pressure is greater than 280 psig.

At least once per 31 days by: I

1. Verifying by flow or position check that each valve (manual, f ower operated, or automatic) in the flow path that is not ocked, sealed, or otherwise secured in position is in its correct position; and Verifying that the isolation valves in the suction line from )

the CST are locked open.

At least once per 18 months by:

Verifying that each motor-driven auxiliary feedwater pump starts automatically, as designed, upon receipt of a test signal and that the respective pressure control valve for each motor-driven pump and each flow control valve with an auto-open feature respond as required; Verifying that the turbine-driven auxiliary feedwater pump starts automatically, as designed, upon receipt of a test signal. The provisions of SpeciAcation 4.0.4 are not applicable for entry into MODE 3; and Verifying that the motor-operated auxiliary feedwater isolation valves and flow control valves close as required upon receipt of an appropriate test signal for steamline differential pressure high coincident with main steam isolation.

SHEARON HARRIS - UNIT 1 3/4 7-5 Amendment No.

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003696185:1

Subject:

Clarification to a Request for Additional Information Related to a License Amendment R equesting a 24-Month Operating Cycle Body:

ADAMS DISTRIBUTION NOTIFICATION.

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t.

1 I

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A001 - OR Submittal: General Distribution Docket: 05000400 Page 2

l FENOC Peny Nuclear Power Plant 10 Center Road Peny, Ohio 44081 John K. liood 440-28&5224 Vice President, Nuclear Fax: 440-28M029 March 20, 2000 PY-CEI/NRR-2479L United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Perry Nuclear Power Plant Docket No. 50-440 Clarification to a Request for Additional Information Related to a License Amendment Requesting a 24-Month Operating Cycle (TAC No. MA5930)

Ladies and Gentlemen:

'On Ja'nuary 17, 2000, the Perry Nuclear Power Plant (PNPP) staff submitted a response to' Request for Additional Information (RAI) associated with a license amendment requesting

. an extension of various surveillance requirements to support a 24-month operating cycle.

During a discussion with the Nuclear Regulatory Commission staff on March 15, 2000, regarding this RAI response, it was determined that two clarifications were desired. The clarifications are:

1. In the response to Question 4, the Surveillance Requirement (SR) being described is SR 3.3.4.1.6 instead of SR 3.3.4.1.

2. In the response to Question 4, the request to extend SR 3.8.4.8 from 60 to 72 months was withdrawn. However, a note in SR 3.8.4.7, which references SR 3.8.4.8 and which was proposed to be modified to reflect the 72 month frequency extension of SR 3.8.4.8, was not revised when the frequency extension fop SR, 3.8.4.8 was withdrawn. Therefore, the note contained in SR 3.8A.7 should remain at its current 60 month frequency.

..There are no regulatory commitments contained in this letter.

If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager - Regulatory Affairs, at (440) 280-5305.

Very truly yours, cc: NRC Project Manager NRC Resident Inspector NRC Region III State of Ohio

4-~(

/ P g c Distri56.txt Distribution Sheet Priority: Normal From: Stefanie Fountain Action Recipients: Copies:

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Subject:

SHEARON HARRIS NUCLEAR POWER PLANT REQUEST FOR LICENSE AMENDMENTL ABORATORY TESTING OF NUCLEAR-GRADE CHARCOAL SUPPLEMENTAL INFORMAT ION Body:

ADAMS DISTRIBUTION NOTIFICATION.

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I Distri56.txt Accession Number ML003693829.

A081 - GL-99-02 MPA L-902, "Laboratory Testing of Nuclear Grade Activated Charcoal" Docket: 05000400 Page 2

Carolina Power & Light Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris NUclear Plant MAR 16 2000 United States Nuclear Regulatory Commission SERIAL: HNP-00-053 ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE CHARCOAL SUPPLEMENTAL INFORMATION

Dear Sir or Madam:

On November 19, 1999, Harris Nuclear Plant (HNP) submitted a proposed license amendment for a revision to the Technical Specifications (TS). The proposed amendment revises the TS to incorporate American Society for Testing and Materials (ASTM) D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," as the standard for testing nuclear-grade activated charcoal. 'pecifically, TS,4.7.6'will be revised for the Control Room Emergency Filtration System, TS 4.7.7 will be revised for the Reactor Auxiliary Building Emergency Exhaust System, and TS 4.9.12 will be revised for the Fuel Handling Building Emergency Exhaust System. These changes are being made in accordance with NRC Generic Letter (GL) 99-02, "Laboratory Testing Of Nuclear-Grade Activated Charcoal," which was issued on June 3, 1999.

In subsequent discussions with the NRC staff, additional information was requested regarding a calculation for flow velocity of applicable charcoal beds for ventilation filtration units. The following information is provided to aid in the review of the proposed change:

Face velocity = design acfm/face area.

40 ft/min = 4000 acfm / 100 ft for the Control Room Emergency Filtration System.

40 ft/min = 6800 acfm / 170 ft for the Reactor Auxiliary Building Emergency Exhaust System.

40 ft/min = 6600-acfm / 165 ft for the Fuel Handling Emergency Exhaust System.

This supplemental information does not affect the conclusions of either the 10 CFR 50.92 evaluation or the Environmental Considerations submitted as part of HNP's November 19, 1999 letter.

COL requests, that the proposed amendment be issued such that implementation;will"'occur within 60 days of issuance to allow time for procedure revision and orderly incorporation, into copies of the Technical Specifications.

lt 5413 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362-2095 (408/

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

Sincerely, J. Scarola, having been first duly swo Q'epose and say that the information contained herein is true and correct to the best of his information, knowledge and belief, and the sources of his information are employees, contractors, and agents of Carolina Power & Light Company.

Notary (Seal)

My commission expires:

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Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Mel Fry, Director, NC DEHNR Mr. R. J. Laufer, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003693268'1

Subject:

REQUEST FOR LICENSE AMENDMENTTO REVISE PROPOSED LICENSE AMENDMENT TO LIMITAPPLICABILITYOF PROVISION TO ALLOWPENETRATIONS TO BE OPEN UN DER ADMINISTRATIVECONTROLS THROUGH CYCLE 10 ONLY.

Body:

ADAMS DISTRIBUTION NOTIFICATION.

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A001 - OR Submittal: General Distribution Docket: 05000400 Page 2

~ 1 Carolina Power 8 tJght Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant MAR 14 2000 United States Nuclear Regulatory Commission SERIAL: HNP-00-059 ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT CONTAINMENTPENETRATIONS

Dear Sir or Madam:

On August 26, 1999, Harris Nuclear Plant (HNP) submitted a proposed license amendment for a revision to the Technical Specifications (TS). The proposed amendment revises the TS 3/4.9.4 to allow containment penetrations to be open, under administrative controls, during core alterations and movement of irradiated fuel assemblies in containment. The proposed amendment was supplemented on February 24, 2000, to specify that administrative controls are not required to maintain radiation dose well below the standard review plan limit with a Containment Building Penetration open during a fuel handling accident in the Containment Building.

The NRC has recently identified a generic industry issue regarding radiation inleakage into the main control room during accident conditions potentially being higher than assumed in certain analysis. Pending resolution of this issue, HNP proposes to revise the proposed license amendment to limit applicability of the provision to allow penetrations to be open under administrative controls through cycle 10 only. This would include refueling outage 9 but would not include refueling outage 10.

This supplemental information does not affect the conclusions of either the 10 CFR 50.92 evaluation or the Environmental Considerations submitted as part of HNP's August 26, 1999 letter.

CP&L requests that the proposed amendment be issued such that implementation will occur within 60 days of issuance to allow time for procedure revision and orderly incorporation into copies of the Technical Specifications.

Please,refer'any questions regarding this submittal, to Mr. J. H. Eads at (919) 362-2646.

Sincerefy,'

54 3 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362.2095 AQG(

J. Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief, and the sources of his information are employees, contractors, and agents of Carolina Power 8r, Light Company.

Notary (Seal)

My commission expires: P Q /00$

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Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Mel Fry, Director, NC DEHNR Mr. R. J. Laufer, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator

ENCLOSURE TO SERIAL: HNP-00-059 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICALSPECIFICATION TS 3/4.9.4 TECHNICALSPECIFICATION PAGES

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REFUEL ING OPERATIONS 3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of four bolts, f'o.po.b<e. of being

b. A minimum of one door in each airlock is closed nd Each penetration providing direct access from the containment atmosphere to e side atmosphere shall be either:

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automatic isolation valve, blind y a m nua flange or equivalent, or

2. Be capable of being closed by OPERABLE automatic normal containment purge and containment pre-entry purge makeup and exhaust isolation valves.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.,

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.

SURVEILLANCE REQUIREMENTS p>b eaa,q~4$ e. o be. 4$ CAoae i Sz saQfI.~~

4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed/isolated conditio or capable of being closed by OPERABLE automatic normal containment purge and containment pre-entry purge makeup and exhaust isolation valves at least once per 7 days during CORE ALTERATIONS or movement of i rradi d fuel in the containment building by: e,haec r <he, hela4

a. Verifying the penetrations are in their closed/isolated condition.

or af'o pd fL ~F Ioesn e (,s isola.k ~

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b. Testing the normal containment purge and containment pre-entry purge makeup and exhaust isolation valves per the applicable portions of Specification 4.6.3.2.

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SHEARON HARRIS - UNIT 1 3/4 9-5 Amendment No. 61

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REFUELING OPERATIONS 3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of four bolts,

b. A minimum of one door in each airlock is capable of being closed*.

and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Be capable of being* closed by a manual or automatic isolation valve, blind flange or equivalent, or

2. Be capable of being closed by OPERABLE automatic normal containment purge and containment pre-entry purge makeup and exhaust isolation valves.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.

SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed/isolated condition, capable of being closed/isolated*, or capable of being closed by OPERABLE automatic normal containment purge and containment pre-entry purge makeup and exhaust isolation valves at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

a. Verifying the penetrations are either closed/isolated or capable of being closed/isolated*, or

b. Testing the normal containment purge and containment pre-entry purge makeup and exhaust isolation valves per the applicable portions of Specification 4.6.3.2.

Penetrations may be opened under administrative controls except for containment purge and exhaust penetrations. This allowance is permitted for refueling outage 9 and cycle 10 only. Operation under these administrative controls has not been approved for refueling outage 10.

SHEARON HARRIS - UNIT 1 3/4 9-5 Amendment No.

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003686510

Subject:

SHEARON HARRIS REQUEST FOR LICENSE AMENDMENTTECHNICAL SPECIFICATIO NS 3/4.9.4 SUPPLEMENTAL INFORMATION Body:

Page 1

~ t, Distri40.txt Docket: 05000400, Notes: Application for permit renewal filed.

Page 2

Carolina Power & Ught Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant SERIAL: HNP-00-021 FEB 34 2000 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS 3/4.9.4 SUPPLEMENTAL INFORMATION

Dear Sir or Madam:

On August 26, 1999, Harris Nuclear Plant (HNP) submitted a proposed license amendment for Technical Specification (TS) TS 3/4.9.4, "Containment Building Penetrations" and associated Bases. HNP has recently revised the Fuel Handling Accident Analysis in the Containment Building which requires modification of the August 26, 1999 submittal. Specifically, the Fuel Handling Accident Analysis demonstrates that administrative controls are not required to maintain radiation dose well below the Standard Review Plan limit with a Containment Building Penetration open during a Fuel Handling Accident in the Containment. Enclosed is the revised TS Bases page incorporating the changes to TS Bases page B3/4 9-1.

This supplemental information does not affect the conclusions of either the 10 CFR 50.92 evaluation or the Environmental Considerations submitted as part of HNP's August 26, 1999 letter.

CPkL requests that the proposed amendment be issued prior to March 31, 2000 to allow implementation prior to HNP Refueling Outage 9.

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

Sincerely, gl)+(y)C 5IO 5413 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362-2095 PQD/

, J. Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief, and the sources of his information are employees, contractors, and agents of Carolina Power &, Light Company.

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Lisa Af Jun.Ja II rO"

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Notary (Seal)

My commission expires: 5" 7.g008 MSE/mse

Enclosures:

1. Summary of Revised Fuel Handling Accident Analysis

2. Technical Specification Bases Page Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Mel Fry, Director, NC DEHNR Mr. R. J. Laufer, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator

, bc:

Ms. D. B. Alexander Mr. M. Janus Mr. G. E. Attarian Mr. C. S. Hinnant Mr. R. H. Bazemore Mr. G. J. Kline Mr. C. L. Burton Mr. Brett Kruse Mr. H. K. Chernoff Ms. Terry Hardy Mr. W. F. Conway Mr. R. D. Martin Mr. G. W. Davis Mr. T. C. Morton Mr. W. J. Dorman Mr. W. S. Orser Mr. R. J. Duncan II Mr. M. Pope Mr. R. J. Field Mr. J. M. Taylor Mr. K. N. Harris Licensing File(s) (2 copies)

Ms. L. N. Hartz Nuclear Records Mr. W. J. Hindman

ENCLOSURE 1 TO SERIAL: HNP-00-021 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICALSPECIFICATION TS 3/4.9.4 SUPPLEMENTAL INFORMATION

SUMMARY

OF REVISED FUEL HANDLINGACCIDENT ANALYSIS On August 26, 1999, Harris Nuclear Plant (HNP) submitted a proposed license amendment for Technical Specification (TS) TS 3/4.9.4, "Containment Building Penetrations" and associated Bases. The previous Fuel Handling Accident Analysis (FHAA), used as the basis for the August 26, 1999 submittal, credited operator action to close penetrations in the event of a fuel handling accident in the Containment Building. Additionally, the previous FHAA did not consider the effect of mixing of the release with the Containment Building atmosphere due to diffusion, convection, and air flow between the Containment Building and the Reactor Auxiliary Building.

The revised FHAA limits breached containment penetrations to penetrations that communicate between the Reactor Containment Building atmosphere and the Reactor Auxiliary Building Ventilation System atmosphere. This is to ensure that a driving force does not exist that would force radioactivity from the containment atmosphere to adjacent atmospheres.

HNP has revised the FHAA to conservatively assume that 30% of the containment free volume, of 2.23E+ 06ft, would mix with the release during a fuel handling accident. This mixing factor of 30% is more conservative than the 50% mixing factor assumed in a similar Technical Specification issued for the Kewaunee Nuclear Power Plant. With Reactor Containment Building Ventilation secured, the air above the reactor cavity will be essentially still, and the released activity would be free to spread in all directions. When the release begins to reach the personnel air lock, (the largest penetration permitted to be open) it will occupy a cylindrical volume with a radius of 65 feet and a height of 90 feet(surface to personnel air lock centerline distance of 45 feet times 2 since the activity also spreads upwards). This calculates to a volume of 1.19E+ 06ft which equals 54% of the containment free volume. Although not credited, if ventilation is in service, mixing could be provided by two safety related fan cooler units that operate to recirculate containment atmosphere at 125,000 ACFM each (HNP Final Safety Analysis Report Table 6.2.2-1).

The revised FHAA demonstrates that doses, at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) during a fuel handling accident in containment, remain well below limits specified in the Standard Review Plan for a release duration of 120 minutes. Additionally, doses to the Control Room staff remain bounded by the Loss of Coolant'Accident Analysis.

The following tables demonstrate that doses remain below the Standard Review Plan limits. The "Pre-Isolation" dose assumes the initial release that equalizes pressure between the Containment Building and the Reactor Auxiliary Building plus the dose due to an assumed conservative release of 500 cfm for 120 minutes. The "During Purge" dose results from exhausting the remaining activity (after the initial 120 minutes) from the Containment Building through a charcoal filter with 90% efficiency for iodine removal.

ENCLOSURE 1 TO SERIAL: HNP-00-021 Pre-Isolation Durin Pur e Pre-isolation+ Pur e Location EAB LPZ EAB LPZ EAB LPZ Wholebod dose rem 6.20E-02 1.41.E-02 6.19E-01 1.40E-01 6.81E-01 1.54E-01 Th roid dose rem 2.36E+01 5.35E+00 2.53E+01 5.73E+00 4.89E+01 1.11E+01

(

The doses for this FHAA compare to the regulatory limits as follows:

Total Doses in This FHAA Regulatory limit This FHAA./limit rem Location EAB LPZ EAB LPZ EAB LPZ Total Whole Bod 6.81E-01 1.54E-01 6.0E+00 6.0E+00 0.114 0.026 Total Th roid 4.89E+01 1.11E+01 7.5E+01 7.5E+01 0.652 0.148 Conservative Assum tions:

The following Conservative assumptions were used in the revised FHAA:

1. A peaking factor of 1.73, instead of the factor of 1.65 in NRC Regulatory Guide 1.25, was

2. The thyroid dose included evaluating thyroid doses.

'I used to determine the fuel activity.

contributed even though this nuclide is not normally used for

3. HNP used a release fraction for of 0.3. The release fraction (0.3) for ' in Regulatory Guide 1.25 is for use in filter sizing. The release fraction normally used for iodines is 0.1.

4. A mixing of 30% of containment volume was assumed.

5. A release flow rate of 500 CFM for 120 minutes was used to calculate off site dose. This results in a total volume of 60,000 ft, even though a volume of 2,427 ft is sufficient to equalize pressure thereby eliminating the driving force to cause radioactivity to exit containment via the open penetration.

Administrative Controls:

To provide additional margin, HNP proposes the following administrative controls for penetrations that are breached during fuel movement in containment or core alterations. HNP proposes to place these administrative controls in TS 3/4.9.4 Bases and plant procedure OMP-003. Future changes to these administrative controls would be in accordance with 10 CFR 50.59.

~ An individual or individuals shall be designated and available at all times, capable of closing the breached penetration.

~ The breached penetrations shall not be obstructed unless capability for rapid removal of obstructions is provided (such as quick disconnects for hoses).

~ For the Personnel Air Lock, at least one door must be capable of being closed.

An additional administrative control is provided to maintain assumptions used in the FHAA:

~ Only penetrations that communicate between the Reactor Containment Building atmosphere and the Reactor Auxiliary Building Ventilation System atmosphere are permitted to be open under these administrative controls.

ENCLOSURE 2 TO SERIAL: HNP-00-021 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICALSPECIFICATION TS 3/4.9.4 SUPPLEMENTAL INFORMATION TECHNICALSPECIFICATION PAGES E2-1

3/4.9 REFUELING OPE ONS

. BASES 3/4.9. 1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses and are specified in the cycle-specific- COLR. The boron concentration limit specified in the COLR ensures that a core K,<< of

< 0.95 is maintained during fuel handling operations. The administrative controls over the required valves during refueling operations precludes the ossibi lity of uncontrolled boron dilution of the filled portion of the RCS.

his action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

Jib@

3/4.9.3 DECAY TIME - DELETED Q~

3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

/4.9.5 COMMUNICATIONS - DELETED Qi SHEARON HARRIS - UNIT 1 B 3/4 9-1 Amendment No. 61

Insett A Penetrations applicable to Technical Specification 3.9.4.b and 3.9.4.c may be opened provided the following administrative controls are in effect:

1. An individual or individuals shall be designated and available at all times, capable of isolating the breached penetration.

2. The breached penetrations shall not be obstructed unless capability for rapid removal of obstructions is provided (such as quick disconnects for hoses).

3. For the Personnel Air Lock, at least one door must be capable of being closed and secured.

4. Only penetrations that communicate between the Reactor Containment Building atmosphere and the Reactor Auxiliary Building Ventilation System atmosphere are permitted to be open under these administrative controls.

Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated, or capable of isolation via administrative controls, on at least one side of containment. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during fuel movement.

3/4.9 REFUELING OPE iONS BASES 3/4.9. 1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS. and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses and are specified in the cycle-specific COLR. The boron concentration limit specified in the COLR ensures that a core K,ff of

< 0.95 is maintained during fuel handling operations. The administrabve controls over the required valves during refueling operations precludes the ossibi lity of uncontrolled boron dilution of the filled portion of the RCS.

his action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. ,

3/4.9.3 DECAY TIME - DELETED 3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization otential while in the REFUELING MODE. Penetrations applicable to Technical pecification 3.9.4.b and 3.9.4.c may be opened provided the following administrative controls are in effect.:

An individual or individuals shall be designated and available at all times, capable of isolating the breached penetration.

The breached penetrations shall not be obstructed unless capability for rapid removal of obstructions is provided (such as quick disconnects for hoses).

For the Personnel Air Lock, at least one door must be capable of being closed and secured.

Only penetrations that communicate between the Reactor Containment Building atmosphere and the Reactor Auxiliary Building Ventilation System atmosphere are permitted to be open under these administrative controls.

Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated, or capable of isolation via adminstratvie controls. on at least one side of containment. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during fuel movement.

3/4.9.5 COMMUNICATIONS - DELETED SHEARON HARRIS - UNIT 1 B 3/4 9-1 Amendment No.

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003685569'1

Subject:

Brunswick, Units 1 and 2,Shearon Harris.Unit 1 and H.B.Robinson- Information related to impact of proposed share exchange transaction.

Body:

Page 1

A L+

Distri50.txt ADAMS DISTRIBUTION NOTIFICATION.

Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003685569.

A001 - OR Submittal: General Distribution Docket: 05000261 Docket: 05000324 Docket: 05000325 Docket: 05000400 Docket: 07200003 Page 2

10 CFR 50.80, 10 CFR 72.$ 0 Carolina Power 8 Light Company PO Box 1551 411 Fayetteville Street Mall Raleigh NC 27602 Serial: PE&RAS00-013 February 14, 2000 U.S. Nuclear Regulatory Commission Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 AND 50-324 / LICENSE NOS. DPR-71 AND DPR-62 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/ LICENSE NO. NPF-63 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / LICENSE NO. DPR-23 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION DOCKET NO. 72-3 / LICENSE NO. SNM-2502 Information related to the im act of the ro osed share exchan e transaction between CP&L Holdin s Iric. and Florida Pro ress Co. orat!on Ladies and Gentlemen:

Pursuant to Section 184 of the Atomic Energy Act, as amended, 10 C.F.R. gf 50.80 and 72.50, and NRC Administrative Letter 96-02, Carolina Power & Light Company (CP&L) submitted an application for transfer of control of its interest in Operating Licenses Nos. DPR-23, DPR-62, DPR-71, and NPF-63 for Brunswick, Units 1 &'2, Robinson, and Harris, and.Materials License No. SNM-2502 for the Robinson Independent Spent Fuel Storage Installation (collectively, CP&L Nuclear Units), by letter dated September 15, 1999. The application requested the consent of the Nuclear Regulatory Commission (NRC) to the indirect transfers of control of CP&L's possessory interest in the licenses of the CP&L Nuclear Units that will occur under the

U.S. Nuclear Regulatory Commission PE&RAS00-013 Page 2 proposed corporate restructuring of CP&L. On December 29, 1999, the NRC issued the Order approving the indirect transfer of control of the subject Operating Licenses from CP&L to CP&L Holdings, Inc. (Holdings), which willbe the new parent company of .

CP&L.

This letter further describes a separate proposed share exchange transaction (Share Exchange) by which Holdings will acquire all of the shares of Florida Progress Corporation (Progress). Progress is the parent company of Florida Power Corporation (FPC), the licensee and operator of the Crystal River Unit 3 nuclear plant (CR-3), as FPC stated in its January 31, 2000 application to the NRC for indirect transfer of control of CR-3 in connection with the Share Exchange.

With the closing of the Share Exchange, Progress will become a direct, wholly-owned subsidiary of Holdings, and FPC will continue to be a wholly-owned subsidiary of Progress and will continue to be the licensee for CR-3. Likewise, CP&L will remain a direct, wholly-owned subsidiary of Holdings and will continue to be the licensee for the

. CP&L Nuclear Units. The CP&L Nuclear Units will not be affected by the Share Exchange with Progress. Holdings will become a registered holding company under the Public Utility Holding Company Act of 1935.

Currently CP&L has 13 members on its Board of Directors. At the time of the internal CP&L restructuring, members of CP&L Board of Directors willbecome members of the Holdings Board of Directors. Upon consummation of the Share Exchange between Holdings and Progress, the number of Holdings directors willbecome 14, 10 of whom will be designated by Holdings and 4 of whom will be designated by Progress, subject to approval by Holdings. Upon consummation of the Share Exchange, all of the members of the Holdings Board willbe U.S. citizens.'oldings will not be owned, controlled, or dominated by an alien, foreign corporation or foreign government.

Holdings stock willbe widely held and publicly traded. Under the terms of the agreement between Holdings and Progress, Progress shareholders will exchange their shares for a combination of cash and Holdings stock. The exact number and distribution of stockholders in Holdings will not be determined definitively until the completion of the Share Exchange.

After'the Share Exchange, CP&L willcon'tinue to be an "electric utility"within .

the meaning of 10 C.F.R. f 50.2, subject to regulation by the North Carolina Utilities Commission, the South Carolina Public Service Commission and the Federal Energy Regulatory Commission.

U.S. Nuclear Regulatory Commission PE& RAS00-013 Page 3 Based on these facts, the Share Exchange between Progress and Holdings does not require any action on the part of the NRC with respect to License Nos. DPR-71, DPR-62, DPR-23, NPF-63 and SNM-2502. No direct or indirect transfer of control of the CP&L NRC licenses, as contemplated by Section 184 of the Atomic Energy Act and 10 C.F.R. g 50.80 or 10 C.F.R. g 72.50, will occur as'a result of the proposed transaction.

There will be no "direct" transfer of control of an NRC license from one legal entity to another, since CP&L will continue to hold the licenses and continue to own its respective interests in the CP&L Nuclear Units upon consummation of the Share Exchange. There also will be no "indirect" transfer of control of the licenses, since CP&L will remain a wholly-owned subsidiary of Holdings, which willcontinue to be the "indirect" owner of the licenses by virtue of its ownership and control of CP&L.

Please call me at 919-546-4579, ifthere are any questions.

Sincerely, ohn R. Caves Regulatory Affairs CC: L. Reyes, Regional Administrator, NRC Region II T. Easlick, Sr'. Resident Inspector, Brunswick, Units 1 and 2 Sr. Resident Inspector, H. B. Robinson J. Brady, Sr. Resident Inspector, Harris Plant A. Hansen, NRR Project Manager, Brunswick; Units 1 and 2 R. Subbaratnam, NRR Project Manager, H. B. Robinson R. Laufer, NRR Project Manager, Harris Plant S. Hom, Esq., OGC

Distri39.txt Distribution Sheet Priority: Normal From: Elaine Walker Action Recipients: Copies:

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NOAC Not Found Total Copies:

Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID'03677619

Subject:

SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NP. 50-400/LICENSE NO. NPF-6 3 REQUEST FOR LICENSE AMENDMENTTECHNICAL SPECIFICATIONS 3/4.2., 3/4.2.3,3/

4.2.5, SUPPLEMENTAL INFORMATION Page 1

Distri39.txt Body:

Docket: 05000400, Notes: Application for permit renewal filed.

Page 2

0

pl CRT Carolina Power & light Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant JAN 19 2000 SERIAL: HNP-00-006 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR L!CENSE AMENDMENT TECHNICALSPECIFICATIONS 3/4.2.2, 3/4.2.3, 3/4.2.5, SUPPLEMENTAL INFORMATION

Dear Sir or Madam:

On July 9, 1999, Harris Nuclear Plant (HNP) submitted a proposed license amendment for Technical Specification (TS) 3/4.2.2, "HEAT FLUX HOT CHANNEL FACTOR FQ(Z)," TS 3/4.2.3, "RCS FLOW RATE AND ENTHALPY RISE HOT CHANNEL FACTOR", TS 3/4.2.5 "DNB PARAMETERS," an associated note in TS Table 2.2-1, and associated Bases. HNP is clarifying requirements for TS 3.2.2 Action a. to be consistent with NUREG-1431, Revision 1.

"Standard Technical Specifications, Westinghouse Plants," dated April 1995. Enclosed is the revised TS page incorporating the changes to TS page 3/4 2-5.

HNP stated in the July 9, 1999 submittal that the changes made were consistent with NUREG-1431, Revision 1. The changes made in this supplemental information are also consistent with NUREG-1431, Revision 1. Therefore, this 'supplemental information does not affect the conclusions of either the 10 CFR 50.92 evaluation or the Environmental Considerations submitted as part of HNP's July', 1999 letter.

CP&L requests that the proposed amendment be issued such that -implementation will occur within 60 days of issuance to allow time for procedure revision and orderly incorporation into copies of the Technical Specifications.

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

Sincerely,

~

F I I / 1 I -I I' ~

P,F, I 'lFII I 'I, il 1

I 5413 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362-2095

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J. Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief, and the sources of his information are employees, contractors, and agents of Carolina Power & Light Company.

qANOpq qfkgP yO

@~ass Notary (Seal)

My commission expires: ( '7 g,D09 MSE/mse

Enclosures:

Technical Specification Page Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Mel Fry, Director, NC DEHNR Mr. R. J. Laufer, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator

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ENCLOSURE TO SERIAL: HNP-00-006 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICALSPECIFICATION TS 3/4.2.2, TS 3/4.2.3, TS 3/4.2.5 TECHNICALSPECIFICATION PAGES

~

POWER DISTRIBUTION L TS 2/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F~Z LIMITING CONDITION FOR OPERATION

~l n e >~< Pcci >c <0 0 3.2.2 FQ(Z) shall be'imited by the following relationships:

Dt:left gg(eK F (Z) ( RTP P

x K(Z) FOR P > 0.5 RTP F

F,(Z) c x K(Z) FOR P ~ 0.5

0.5 Where

RTP F0 the FQ limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106, P = THERMAL POWER , and RATED THERMAL POWER K(Z) = the normalized F.(Z) as a function of core height specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With F~(Z) exceeding its limit:

Reduce THERMAL POWER at least 1X for each lX FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Po er Range Neutron Flux-High Trip Setpoints within the next hours; POWER OPERATION m'ay proceed for up to a total o 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the ~@

Overpower bT Trip Setpoints have been reduced at least 1X for each 1R F~(Z) exceeds the limit.

\

Identify and correct the cause ot the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limit.

Othc~lse, Ls in a4 lens% PiDpE zvlMii lo .~iiu~>

SHEARON HARRIS - UNIT 1 3/4 2-5 Amendment No. 25

POWER DISTRIBUTION L TS

> ' 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F~~Z LIHITING CONDITION FOR OPERATION 3.2.2 F~(Z) shall be within the limits specified in the COLR.

APPLICABILITY: NODE 1.

ACTION:

With FQ(Z) exceeding its limit:

Reduce THERMAL POWER at least 1X for each 1X F~(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower dT Trip Setpoints have been reduced at least 1X for each lt Fo(Z) exceeds the limit. Otherwise, be in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERNL POWER may then be increased provided F,(Z) is demonstrated through incore mapping to be within its limit.

SHEARON HARRIS - UNIT 1 3/4 2-5 'mendment No.

]AN 2 S 5N 4

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)////~O Distri-8.txt Distribution Sheet Priority: Normal From: Elaine Walker Action Recipients: Copies:

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Internal ecipieiits:

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'idsManager OK RidsAcrsAcnwMailCenter 1 OK OGC/RP Not Found OGC Not Found NRR/DSSA/SRXB Not Found N RR/DSSA/SPLB Not Found ACRS Not Found External Recipients:

NOAC Not Found Total Copies:

Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID'03676878

Subject:

REVISED TECHNICAL SPECIFICATION PAGES FOR LICENSE AMENDMENTREQUEST-ADDITION OF METHODOLOGY REFERENCES TO CORE OPERATING REPORT Body:

Page 1

Distri-8.txt Docket: 05000400, Notes: Application for permit renewal filed.

Page 2

Carolina Power & Ught Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant SERIAL: HNP-00-003 JAN 11 2000 10 CFR 50.90 United States Nuclear Regulatory Commission ATI'ENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REVISED TECHNICALSPECIFICATION PAGES FOR LICENSE AMENDMENT REQUEST - ADDITIONOF METHODOLOGY REFERENCES TO CORE OPERATING LIMITS REPORT

Dear Sir or Madam:

By letter dated August 4, 1999, Carolina Power & Light Company (CP&L) requested a revision to the Technical Specifications (TS) for the Harris Nuclear Plant (HNP) to incorporate analytical methodology references in TS 6.9.1.6.2 which are used to determine core operating limits. These analytical methodologies are documented in topical reports which have been accepted by the Nuclear Regulatory Commission (NRC) for referencing in licensing applications. By letter dated December 3, 1999, CP&L submitted a re-typed TS page.

CP&L has revised the affected TS pages for the subject license amendment request to incorporate the format for referencing approved Siemens Power Corporation topical reports in the TS. The specific revision number and date of the referenced topical reports are specified in the Core Operating Limits Report (COLR).

These changes do not affect the conclusions of either the 10CFR50.92 or the Environmental Considerations evaluations previously submitted.

Enclosure 1 provides page change instructions for incorporating the proposed revisions.

Enclosure 2 provides the proposed TS pages.

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

5413 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362-2095 ALQO~g( p7'5

Document Control Des SERIAL: HNP-00-002 Page 2 Sincerely, AEC

Enclosures:

1. Page Change Instructions

2. Technical Specification Pages James Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are employees, contractors, and agents of Carolina Power & Light Company.

Notary (Seal)

>ggltlt tl~

My commission expires: pj. lgq ">i~

0 ~i f~~ gOIA Ry Q~g Mr. J. B. Brady, NRC Sr. Resident Inspector

<on as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," approved version as specified in the COLR.

(Methodology for Specification 3.2. 1 - Axial Flux Difference, 3.2.2 - Heat, Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

XN-NF-78-44(P)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR.

(Methodology for Specification 3. 1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6'-'24a Amendment No.

ADMINISTRATIVE CONTRO 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

ANF-88-054(P)(A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H'. B. Robinson Unit 2," approved version as specified in the COLR.

(Methodology for Specification 3.2. 1 - Axial Flux Difference, and 3.2.2 - Heat. Flux Hot Channel Factor).

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specification 3. 1. 1.2 - SHUTDOWN MARGIN - MODES 3, 4 AND 5, 3.2.2 - Heat Flux Hot Channel Factor and 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor).

WCAP-10266-P-A. Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE", March 1987 (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

WCAP-11837-P-A. "EXTENSION OF METHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES", January 1990 (W Proprietary).

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-081(P)(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.

(Methodology for Specification 3. 1. 1.3 - Moderator Temperature Coefficient., 3. 1.3.5 Shutdown Bank Insertion Limits, 3. 1.3.6-Control Bank Insertion Limits, 3.2. 1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approve version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

n. XN-NF-82-49(P)(A), "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," approved version as specified in the COLR.

(Methodology for Specification 3.2. 1 - Axial Flux Difference.

3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24b Amendment No.

C ADMINISTRATIVE CONTROL r

6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

o. EMF-96-029(P)(A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.

for Specification 3. 1. 1.2 - SHUTDOWN MARGIN - MODES 3,

'Methodology 4 and 5, 3.1. 1.3 - Moderator Temperature Coefficient, 3. 1.3.5-Shutdown Bank Insertion Limits, 3. 1.3.6 - Control Bank Insertion Limits. 3.2. 1 - Axial Flux Difference. 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9. 1 - Boron Concentration).

6.9. 1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9. 1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10CFR50.4 within the time period specified for each report.

6~10 DELETED (PAGE 6-25 DELETED)

SHEARON HARRIS - UNIT 1 6-24c Amendment No.

JAN 24 2' Distri66.txt Distribution Sheet Priority: Normal From: Esperanza Lomosbog Action Recipients: Copies:

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NRC PDR Not Found NOAC Not Found Total Copies: 19 Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 993470195

Subject:

SHEARON HARRIS NUCLEAR POWER PLANT RETYPED TECHNICAL SPECIFICATION PAG E FOR ADDITION OF METHODOLOG REFERENCES TO CORE OPERATING LIMITS REPO RT Body:

PDR ADOCK 05000400 P Docket: 05000400, Notes: Application for permit renewal filed.

Page 1

f Distri66.txt Page 2

CAGE Carolina Power & Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 SERIAL: HNP-99-179 GEO S'I989 10 CFR 50.90 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RETYPED TECHNICALSPECIFICATION PAGE FOR ADDITIONOF METHODOLOGY REFERENCES TO CORE OPERATING LIMITSREPORT

Dear Sir or Madam:

On August 4, 1999, Carolina Power & Light Company (CP&L) requested a License Amendment to revise the Harris Nuclear Plant (HNP) Technical Specification (TS) 6.9.1.6.2 related to "Methodology References to Core Operating Limits Report." As a result of a recently issued License Amendment, CP&L is submitting a retyped TS page which supersedes a page included in the August 4, 1999 submittal.

On October 19, 1999, the NRC issued License Amendment No. 92 to the HNP TS, which revised multiple TS pages as a result of changes to TS Section 6.5, "Review and Audit;" TS 6.8.2, TS'1 and TS Section 6.10, "Record Retention." Issuance of Amendment 92 independently '.8.3, affected page 6-24c from the August 4, 1999 request, which is currently under NRC review. contains the retyped TS page which incorporates changes resulting from the issuance of Amendment No. 92.

Please refer any questions regarding this matter to Mr. J. H. Eads at (919) 362-2646.

incerely, Donna B. Alexander Manager, Regulatory Affairs Harris Nuclear. Plant 'l

'I 0 I 4

AEC/twk 9,'s~ 70ll'5'.""' ".' '::

c: Mr. J. B. Brady (NRC Senior Resident Inspector, HNP)

R. J. Laufer (NRR Project Manager, HNP)

'r.

Mr. M. Fry (Director N.C. DEHNR)

Mr. L. A. Reyes (NRC Regional Administrator, Region II) 5413 Shearon Harris Road New Hill NC

ENCLOSURE 0 SERIAL: HNP-99-179 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RETYPED TECHNICALSPECIFICATION PAGE FOR ADDITIONOF METHODOLOGY REFERENCES TO CORE OPERATING LIMITS REPORT RETYPED TS PAGE 6-24c

I

.v ADMINISTRATIVE CONTROL 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

(Methodology for Specification 3.2. 1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor ).

o. EMF-96-029(A), Volume 1, Volume 2 and Attachment, "Reactor Analysis Systems for PWRs," Siemens Power Corporation, Richland WA 99352.

(Methodology for Specification 3. 1. 1.2 SHUTDOWN MARGIN - MODES 3, 4 and 5, 3. 1. 1.3 - Moderator Temperature Coefficient, 3. 1.3.5-Shutdown Bank Insertion Limits. 3. 1.3.6 - Control Bank Insertion Limits, 3.2. 1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channnel Factor, and 3.9. 1 - Boron Concentration.)

6.9. 1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9. 1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10CFR50.4 within the time period specified for each report.

6. 10 DELETED (PAGE 6-25 DELETED)

SHEARON HARRIS - UNIT 1 6-24c Amendment No.

Distri55.txt Distribution Sheet Priority: Normal From: Esperanza Lomosbog Action Recipients: Copies:

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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 993310058

Subject:

SHEARON HARRIS NUCLEAR POWER PLANTDOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORY TESTING OF NUCLEAR-GRADE ACT IVATED CHARCOAL Body:

pdr adock 05000400 p Docket: 05000400, Notes: Application for permit renewal filed.

Page 1

Carolina Power 8 Light Company lames Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant SERIAL: HNP-99-166 NOV 19 1999 10 CFR 50.90 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL

Dear Sir or Madam:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power 0 Light Company (CPS') requests a revision to the Technical Specifications (TS) for the Harris Nuclear Plant (HNP). The proposed amendment revises the TS to incorporate American Society for Testing and Materials (ASTM) D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," as the standard for testing nuclear-grade activated charcoal. Specifically, TS 4.7.6 will be revised for the Control Room Emergency Filtration System, TS 4.7.7 will be revised for the Reactor Auxiliary Building Emergency Exhaust System, and TS 4.9.12 will be revised for the Fuel Handling Building Emergency Exhaust System. These changes are being made in accordance with NRC Generic Letter (GL) 99-02, "Laboratory Testing Of Nuclear-Grade Activated Charcoal," which was issued on June 3, 1999. The response to GL 99-02 for HNP is being mailed separately. provides a description of the proposed changes and the basis for the changes. details, in accordance with 10 CFR 50.91(a), the basis for CP8cL's determination that the proposed changes do not involve a significant hazards consideration. provides an environmental evaluation which demonstrates that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment is required for approval of this amendment request. provides page change instructions for incorporating the proposed revisions.

r provides the proposed TS pages.

~ ~

s Document Control Desk SERIAL: HNP-99-166 Page 2 CP&L requests that the proposed amendment be approved by January 31, 2000 to avoid adversely impacting scheduled charcoal testing. CP&L also requests that the proposed amendment be issued such that implementation will occur within 60 days of issuance to allow time for orderly incorporation into copies of the TS.

In accordance with 10 CFR 50.91(b), CP&L is providing the State of North Carolina a copy of this license amendment request.

Please refer any questions regarding this submittal to Mr. J. H. Eads at (919) 362-2646.

Sincerely, nA)

AEC

Enclosures:

1. Basis for Change Request

2. 10 CFR 50.92 Evaluation

3. Environmental Considerations

4. Page Change Instructions

5. Technical Specification Pages James Scarola, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are employees, contractors, and agents of Carolina Power & Light Company.

Notary (Sea

~g~~ltu p+~ S. >4~

My commission expires:

P f gOTARy O~

g~~ Ore CC: Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. Mel Fry, Director, N.C. DEHNR e ~~ y'UBLi<

Mr. R. J. Laufer, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator iiiiliillllli to SERIAL HNP-99-166 Page 1 of 3 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL BASIS FOR CHANGE RE UEST

~Back round NRC Generic Letter (GL) 99-02, "Laboratory Testing Of Nuclear-Grade Activated Charcoal,"

was issued on June 3, 1999. GL 99-02 alerted the nuclear power industry that testing nuclear-grade activated charcoal to standards other than American Society for Testing and Materials (ASTM) D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," is not acceptable to the NRC. A series of laboratory testing has demonstrated that test standards other than ASTM D3803-1989 do not provide accurate and reproducible results. These other test standards may overestimate the capability of the charcoal to adsorb radioiodine.

Analyses of design-basis accidents assume a particular engineered safety feature (ESF) charcoal filter adsorption efficiency when calculating offsite and control room operator doses. Charcoal filter samples are tested to determine whether the filter adsorption efficiency is greater than that assumed in the design-basis accident analysis. Accurate charcoal testing confirms the capability of the charcoal filters in ESF ventilation systems to adsorb radioiodine such that the dose limits of General Design Criterion 19 and Part 100 of the Code of Federal Regulations are not exceeded. The NRC adopted ASTM D3803-1989 as an acceptable testing standard since it is the only available testing standard that provides accurate and reproducible test results.

The Technical Specification (TS) ESF ventilation systems at the Harris Nuclear Plant (HNP) are the Control Room Emergency Filtration System, the Reactor AuxiliaryBuilding Emergency Exhaust System and the Fuel Handling Building Emergency Exhaust System. The TS Surveillance Requirements for these systems currently specify that laboratory analyses of representative carbon samples meet the testing criteria of Regulatory Position C.6.a of Regulatory Guide (RG) 1.52, Revision 2 (March 1978), "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants." A methyl iodide penetration limit is specified for the testing parameters of 30'C and at a relative humidity of 70%

in accordance with ASTM D3803 (no year specified). The 1989 version of ASTM D3803 is not currently referenced as the testing standard in HNP TS. The plant procedure governing this testing currently specifies ASTM D3803-1979 as the testing protocol.

Pro osed Chan es Technical Specification Surveillance Requirements 4.7.6.b.2 and 4.7.6.c for Control Room Emergency Filtration System (CREFS) operability, 4.7.7.b.2 and 4.7.7.c for Reactor Auxiliary Building Emergency Exhaust System (RABEES) operability, and 4.9.12.b.2 and 4.9.12.c for Fuel Handling Building Emergency Exhaust System (FHBEES) operability will be revised to incorporate the ASTM D3803-1989 standard.

~ ~

J Enclosure 1 to SERIAL HNP-99-166 Page 2 of 3 Basis ASTM D3803-1989 has two additional testing periods that are not required by other standards:

the stabilization period and the equilibration period. During the stabilization period, the charcoal bed is brought to thermal equilibrium with the test temperature before the start of pre-equilibration. During the equilibration period, air at the test temperature and relative humidity (RH) is passed through the charcoal beds to ensure the charcoal adsorbs all the available moisture before the feed period. During this period, the system is more closely monitored than in the pre-equilibration period to ensure that all parameters are maintained within their limits. The major elements of the ASTM D3803-1989 test are as follows:

95% RH or 70% RH (for ESF systems that control the RH to 70% or less) 2-hour minimum thermal stabilization, at 30 'C [86 'F]

16-hour pre-equilibration (pre-sweep) time, with air at 30 'C [86 'F] and plant-specific RH 2-hour equilibration time, with air at 30 'C [86 'F] and plant-specific RH 1-hour challenge, with gas at 30 'C [86 'F] and plant-specific RH 1-hour elution (post-sweep) time, with air at 30 'C [86 'F] and plant-specific RH ASTM D3803-1989 is more stringent than other testing standards because it does not differentiate between new and used charcoal. It has a longer equilibration period performed at a temperature of 30 'C [86 'F] and an RH of 95% (or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test.

The temperature and relative humidity at which the charcoal samples will be tested are 30' (86') and 70%, respectively. These parameters remain unchanged from the present HNP TS requirements. The acceptable methyl iodide penetration specified in the TS will be changed, as calculated using the following formula provided in GL 99-02:

100% - Meth 1 Iodide Efficienc for Charcoal Credited in Accident Anal sis Safety Factor (> 2)

The HNP Final Safety Analysis Report (FSAR) assumes a 99% efficiency for CREFs, a 95%

efficiency for RABEES and a 95% efficiency for FHBEES. A safety factor = 2 is utilized.

Revising the HNP TS to incorporate ASTM D3803-1989 as the testing standard for nuclear-grade activated charcoal samples will ensure that accurate charcoal testing confirms the capability of the charcoal filters to adsorb radioiodine. This, in turn, will ensure that the ESF ventilation systems at HNP are capable of performing their safety function of reducing the potential onsite and offsite consequences of a radiological accident.

to SERIAL HNP-99-166 Page 3 of 3 The next laboratory surveillance tests of TS ESF ventilation system charcoal samples are currently scheduled for the first quarter of 2000. These samples are to be tested within 31 days from the removal date, and will be tested in accordance with ASTM D3803-1989. Charcoal samples in the CREFS, FHBEES and RABEES will continue to be tested in accordance with ASTM D3803-1989, in lieu of current TS-required laboratory testing, until this license amendment is approved by the NRC. As stated in GL 99-02, the NRC will exercise enforcement discretion to eliminate unnecessary testing of charcoal samples to both ASTM D3803-1989 and the current TS testing protocol during the period of time between issuance of GL 99-02 and approval of this TS amendment.

to SERIAL HNP-99-166 Page1of 2 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL 10 CFR 50.92 EVALUATION The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Carolina Power & Light Company has reviewed this proposed license amendment request and determined that its adoption would not involve a significant hazards determination. The basis for this determination is provided below.

Pro osed Chan es Technical Specification Surveillance Requirements 4.7.6.b.2 and 4.7.6.c for Control Room Emergency Filtration System (CREFS) operability, 4.7.7.b.2 and 4.7.7.c for Reactor Auxiliary Building Emergency Exhaust System (RABEES) operability, and 4.9.12.b.2 and 4.9.12.c for Fuel Handling Building Emergency Exhaust System (FHBEES) operability will be revised to incorporate the ASTM D3803-1989 standard.

Basis The changes do not involve a significant hazards consideration for the following reasons:

The proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

This proposed change to revise the standard to which activated charcoal samples are tested will ensure that testing is accurate and repeatable. This will help ensure that the Engineered Safety Feature (ESF) ventilation systems are capable of performing their safety function. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes incorporate ASTM D3803-1989 as the testing standard for nuclear-grade activated charcoal samples. This will ensure that testing is accurate and repeatable. Plant structures, systems, and components will not be operated in a different manner as a result of these proposed changes and no physical modifications to equipment are involved. Using the improved testing protocol does not have the potential for creating the possibility of a new or different type of accident from any previously evaluated.

to SERIAL HNP-99-166 Page 2 of 2

3. The proposed amendment does not involve a significant reduction in the margin of safety.

The proposed changes do not change the manner in which structures, systems or components are operated. Revising the standard to which activated charcoal samples are tested will ensure that testing is accurate and repeatable. This will help ensure that the ESF ventilation systems are capable of performing their safety function. Therefore, the proposed changes do not involve a reduction in the margin of safety.

Conclusion Based on the above evaluation, it is concluded that the proposed amendment does not involve a significant hazards consideration.

to SERIAL HNP-99-166 Page1of 2 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL ENVIRONMENTALCONSIDERATIONS 10 CFR 51.22(c)(9) provides criteria for licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license'or a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (3) result in a significant increase in individual or cumulative occupational radiation exposure. Carolina Power 4 Light Company has reviewed this request and determined that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination is provided below.

Pro osed Chan es Technical Specification Surveillance Requirements 4.7.6.b.2 and 4.7.6.c for Control Room Emergency Filtration System (CREFS) operability, 4.7.7.b.2 and 4.7.7.c for Reactor Auxiliary Building Emergency Exhaust System (RABEES) operability, and 4.9.12.b.2 and 4.9.12.c for Fuel Handling Building Emergency Exhaust System (FHBEES) operability will be revised to incorporate the ASTM D3803-1989 standard.

Basis The change meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in Enclosure 2, the proposed amendment does not involve a significant hazards consideration.

2. The proposed amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed changes do not involve any new equipment or require existing systems to perform a different type of function than they are currently designed to perform. The proposed change to revise the standard to which activated charcoal samples are tested will ensure that testing is accurate and repeatable. This will help ensure that the Engineered Safety Feature (ESF) ventilation systems are capable of performing their safety function.

The changes do not introduce any new effluents or increase the quantities of existing effluents. As such, the changes cannot affect the types or amounts of any effluents that may be released offsite.

0 to SERIAL HNP-99-166 Page 2 of 2

3. The proposed amendment does not result in an increase in individual or cumulative occupational radiation exposure.

The proposed change does not result in any physical plant changes or new surveillance which would require additional personnel entry into radiation controlled areas.

Therefore, the amendment will not result in an increase in individual or cumulative occupational radiation exposure.

to SERIAL HNP-99-166 Page1of 1 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63

- REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL PAGE CHANGE INSTRUCTIONS 3/4 7-15 3/4 7-15 3/4 7-17 3/4 7-17 3/4 7-18 3/4,7-18 3/4 9-15 3/4 9-15

0 to SERIAL HNP-99-166 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT LABORATORYTESTING OF NUCLEAR-GRADE ACTIVATEDCHARCOAL TECHNICALSPECIFICATION PAGES

PLANT SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM

'SURVEILLANCE REQUIREMENTS (Continued)

P<~

. Revision92. March 1978, and the system flow rate is 4000 cfm + 10K during system operation when tested in accordance with ANSI N510-1980; and

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 Regulat Guide 1.52, Revision 2, March 1978 mee s the a oratory tes sng cn ena o egu a ory Position C.6.a of R ulatory Guide 2. Revision 2 March 1978. b showin a meth l ao ide penetra ion o ess an . w en este at a temperature of 30'C an a a re a i e hum> ity of 70X in accordance with ASTM D380 (qgq Dd~ 0 5 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 Positio C.6.b of Re ulator Guide 1.52

'on 2. March 1978 mee s e aboratory es ing criteria of Regulatory os> ion .6 a o r Guid 2 vision 2, March 19 8, b showin a methyl iodide penetration of less than

0. 175 hen tested at a temperature of 30'C and at a relative umidity of 70K in cordance with ASTM D3803.

Aha At least once per 18 months by:

-RP

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5. 1 inches water gauge while operating the system at a flow rate of 4000 cfm + 10K;

2. Verifying that, on either a Safety Injection or a High Radiation test signal, the system automatically switches into an isolation with recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks; Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch Mater Gauge at less than or equal to a pressurization flow of 315 cfm relative to adjacent areas during system operation;

4. Verifying that the heaters dissipate 14 + 1.4 kW when tested in accordance with ANSI N510-1980; and ~c.W

5. Deleted.

Zg.4+

SHEARON HARRIS - UNIT 1 3/4 7-15 Amendment No. 10

PLANT SYSTEMS 3/4.7. 7 REACTOR AUXILIARY BUILDING RAB EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent RAB Emergency Exhaust Systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one RAB Emergency Exhaust 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 4.7.7 Each RAB Emergency Exhaust System shall be demonstrated OPERABLE:

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 verifying that the system operates for at least 10 continuous hours with the heaters operating; At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings. or (2) following significant painting, fire, or chemical release in any ventilation zone communicating with the system by:

Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of'ess than 0.05K and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the unit flow rate is 6800 cfm + 10K during system operation when tested in accordance with ANSI N510-1980;

2. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in >~(e4 accordance with Regulatory Position C.6.b of Re ul Guide 1.52, Revision March 1978 mee s the laboratory es ing cr~ erma of Regu atory osition C.6.a of Re ulator Guide 1.52, Revisi n , March 1978. b owin me yl so i e penetration of less than 1.0K when teste a a temperature oj 30'C an a a re a ive humidity of 70K in accordance with A 3803 a ggog @ -(Rl l AQ After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months o c arcoal 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 'evision 2, March 1978, SHEARON HARRIS - UNIT 1 3/4 7-17 g~~$ Mo,

PLANT SYSTEMS REACTOR AUXILIARY BUILDING RAB EMERGENCY EXHAUST SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Pe.H meets the laboratory testing criteria of Regulatory Position' C. .a of Re ulator Guid 1.52, ision 2, March 1978, b s met y iodide penetra ion of less than 1.0 when ested at a temperature of 30'C and at a re a sve umidity of 70K in accordance with ASTM D3803

-ter~ 4 gS'I Qh At least once per 18 months y: ~

1. Verifying that the pressure drop across the combined HEPA filters and charcoa'1 adsorber bank is less than 4. 1 inches water gauge while operating the unit at a flow rate cfm + lOX.

of'800

2. Verifying that the system starts on a Safety Injection test signal, Verifying that the system maintains the areas served by the exhaust system at a negative pressure of greater than or equal to 1/8 inch water gauge relative to the outside atmosphere,

4. Verifying that the filter cooling bypass valve is locked in the balanced position, and

5. Verifying that the heaters dissipate 40 + 4 kW when tested in accordance with ANSI N510-1980.

After each complete or partial replacement of a HEPA filter bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a DOP test aerosol while operating the unit at a flow rate of 6800 cfm + 10K; and After each complete or partial replacement of a charcoal adsorber bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the unit at a flow rate of 6800 cfm + 10K.

SHEARON HARRIS - UNIT 1 3/4 7-18 Ae rain> ~~

REFUELING OPERATIONS FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.9.12 (Continued)

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 Re ula Guide 1.52, Revision 2, March 1978. mee s t e aboratory es ing cra crea o egu atory osition C.6.a o Re ulatory Guide 1.52. Rev' March 1978, b showin a met yl io i e penetration o less than 1.0X when teste a a temperature of 30'C an a re a ave humidity of 70'n accordance with AST 0

-l08'I ~>~

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of c arcoa adsorber operation by verifying, within 31 days after removal. that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory P ition C.6.b of Regulatory Guide 1.52 'evision h 1978 meets t e a ora ory es ~ng cri eria o egu atory Position C.6.a of Re ulator Guide 1 2 evision 2, March 1978 b showin me yl iodide penetration of s than 1.0X w en este a a temperature of 30'C and at a relative umph )ty of 70K in ccordance with ASTM 03803 has e hh, -ln8e At east once per 18 months by: Aaa Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber bank is not greater than 4. 1 inches water gauge while operating the unit at a flow rate of 6600 cfm + 10K.

Verifying that. on a High Radiation test signal, the system automatically starts and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/8 inch water gauge, relative to the outside atmosphere.

during system operation at a flow rate of 6600 cfm + 10K,and

4. Deleted Verifying that the heaters dissipate 40 + 4 kW when tested in accordance with ANSI N510-1980.

After each complete or partial replacement of a HEPA filter banks by verifying that the unit satisfies the in-place penetr ation leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for- a DOP test aerosol while operating the unit at a flow rate of 6600 cfm + 10K.

SHEARON HARRIS - UNIT 1 3/4 9-15 Amendment No. 82

V PLANT SYSTEMS CONTROL ROOM EMERGENCY FILTRATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Revision 2, March 1978, and the system flow rate is 4000 cfm f

+ 10K during system oper ation when tested in accordance with ANSI N510-1980; and 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, has a methyl iodide penetration of ( 0.5X when tested at a temperature of 30'C and at a relative humidity of 70K in accordance with ASTM D3803-1989. I 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, March 1978, has a methyl iodide penetration of ( 0.5X when tested at a temperature of 30 C and at a relative humidity of 70K in accordance with ASTM 03803-1989. I At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5. 1 inches water gauge while operating the system at a flow rate of 4000 cfm + 10K;

2. Verifying that, on either a Safety Injection or a High Radiation test signal, the system automatically switches into an isolation with recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks; Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 315 cfm relative to adjacent areas during system operation;

4. Verifying that the heaters dissipate 14 + 1.4 kW when tested in accordance with ANSI N510-1980; and

5. Deleted.

SHEARON HARRIS - UNIT 1 3/4 7-15 Amendment No.

PLANT SYSTEMS 3/4.7. 7 REACTOR AUXILIARY BUILDING RAB EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent RAB Emergency Exhaust Systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one RAB Emergency Exhaust 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 4.7.7 Each RAB Emergency Exhaust 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 verifying that the system operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following significant painting, fire, or chemical release in any ventilation zone communicating with the system by:

Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05K and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978. and the unit flow rate is 6800 cfm + 10K during system operation when tested in accordance with ANSI N510-1980;

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, has a methyl iodide penetration of ( 2.5X when tested at a temperature of 30'C and at a relative humidity of 70K in accordance with ASTM D3803-1989. I

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, March 1978.

C SHEARON HARRIS - UNIT 1 3/4 7-17 Amendment No.

PLANT SYSTEMS REACTOR AUXILIARY BUILDING RAB EMERGENCY EXHAUST SYSTEM SURVEILLANCE REQUIREMENTS (Continued) has a methyl iodide penetration of < 2.5X when tested at a temperature of 30'C and at a relative humidity of 70K in accordance with ASTH D3803-1989.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber bank is less than 4. 1 inches water gauge while operating the unit at a flow rate of 6800 cfm + 10K.

Verifying that the system starts on a Safety Injection test signal, Verifying that the system maintains the areas served by the exhaust system at a negative pressure of greater than or equal to 1/8 inch water gauge relative to the outside atmosphere.

4. Verifying that the filter cooling bypass valve is locked in the balanced position, and

5. Verifying that the heaters dissipate 40 + 4 kW when tested in accordance with ANSI N510-1980.

e. After each complete or partial replacement of a HEPA filter bank.

by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a DOP test aerosol while operating the unit at a flow rate of 6800 cfm + 10K; and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the unit at a flow rate of 6800 cfm + 10K.

SHEARON HARRIS - UNIT 1 3/4 7-18 Amendment No.

4 1

I A

V

~a o

REFUELING OPERATIONS FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.9.12 (Continued)

Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b'f Regulatory 1.52, Revision 2, March 1978, has a methyl iodide 'uide penetration of < 2.5X when tested at a temperature of 30'C and at a relative humidity of 70K in accordance with ASTM D3803-1989. I 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.

March 1978, has a methyl iodide penetration of < 2.5X when tested at a temperature of 30 C and at a relative humidity of 70K in accordance with ASTM D3803-1989. I

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber bank is not greater than 4.1 inches water gauge while operating the unit at a flow rate of 6600 cfm + 10K,

2. Verifying that, on a High Radiation test signal, the system automatically starts and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/8 inch water gauge, relative to the outside atmosphere, during system operation at a flow rate of'600 cfm + 10K,and 4 Deleted

5. Verifying that the heaters dissipate 40 + 4 kW when tested in accordance with ANSI N510-1980.

After each complete or partial replacement of a HEPA filter bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a DOP test aerosol while operating the unit at a flow rate of 6600 cfm + 10K.

SHEARON HARRIS - UNIT 1 3/4 9-15 Amendment No.

Distri42.txt Distribution Sheet Priority: Normal From: Andy Hoy Action Recipients: Copies:

R Laufer 1 Not Found NRR/DLPM/LPD2-2 1 Not Found E Dunnington 1 Not Found Internal Recipients:

RidsManager OK OGC/RP Not Found NRR/DSSA/SRXB 'Not Found NRR/DSS SPLB Not Found ie Center 8 Not Not Found Found External Recipients:

NRC PDR Not Found NOAC Not Found Total Copies:

Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 993160242

Subject:

RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ALTER NATIVE PLAN FOR SPENT FUEL POOLS C & D COOLING AND CLEANUP SYSTEM P IPING Body:

Docket: 05000400, Notes: Application for permit renewal filed.

Page 1

~ 4 Carolina Power & Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 SERIAL: HNP-99-172 OCT 39 1999 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATIONREGARDING THE ALTERNATIVE PLAN FOR SPENT FUEL POOLS C & D COOLING AND CLEANUP SYSTEM PIPING

Dear Sir or Madam:

By letter HNP-98-188, dated December 23, 1998, Carolina Power & Light Company (CP&L) submitted a license amendment request to increase fuel storage capacity at the Harris Nuclear Plant (HNP) by placing spent fuel pools C & D in service. The U. S. Nuclear Regulatory Commission (NRC) issued letters dated March 24, 1999, April 29, 1999, June 16, 1999, and August 5, 1999 requesting additional information regarding our license amendment application.

HNP letters HNP-99-069, dated April 30, 1999, HNP-99-094, dated June 14, 1999, HNP-99-112, dated July 23, 1999, and HNP-99-129, dated September 3, 1999 provided our respective responses.

By letter dated September 20, 1999, the NRC issued a fifth request for additional information (RAI) regarding our license amendment application to place spent fuel pools C & D in service.

The September 20, 1999 NRC RAI specifically requests additional information on the proposed alternative plan to demonstrate compliance with ASME Code requirements for the cooling and cleanup system piping in accordance with 10 CFR 50.55a(a)(3)(i). The Enclosures to this letter provide the HNP response to the NRC staff's September 20, 1999 RAI.

The enclosed information is provided as supplement to our December 23, 1998 amendment request and does not change our initial determination that the proposed license amendment represents a no significant hazards consideration.

5413 Shearon Harris Road New Hill NC

Document Control Desk SERIAL: HNP-99-172 Page 2 Please refer any questions regarding the enclosed information to Mr. Steven Edwards at (919) 362-2498.

incerely, Donna B. Alexander Manager, Regulatory Affairs Harris Nuclear Plant KWS/kws

Enclosures:

1. HNP Responses to NRC Request For Additional Information (RAI)

2. Technical Report: HNP - Material Identification of Chips from Carbon Steel Welds Associated with the Spent Fuel Pool Activation Project (1 page total)

3. Chemistry Sample Data Sheets (2 sheets total)

4. QCI-19.1, Revision 1, entitled "Preparation Ec Submittal of Weld Data Report, Repair Weld Data Report, Tank Fabrication Weld Record A Seismic I Weld Data Report" (25 pages total)

Mr. J. B. Brady, NRC Senior Resident Inspector (w/ Enclosure 1)

Mr. Mel Fry, N.C. DEHNR (w/ Enclosure 1)

Mr. R. J. Laufer, NRC Project Manager (w/ all Enclosures)

Mr. L. A. Reyes, NRC Regional Administrator - Region II (w/ Enclosure 1)

Document Control Desk SERIAL: HNP-99-172 Page 3 bc: (all w/Enclosure 1)

Mr. K. B. Altman Ms. L. N. Hartz Mr. G. E. Attarian Mr. W J. Hindman Mr. R. H. Bazemore Mr. C. S. Hinnant Mr. C. L. Burton Mr. W. D. Johnson Mr. S. R. Carr Mr. G. J. Kline Mr. J. R. Caves Mr. B. A. Kruse Mr. H. K. Chernoff (RNP) Ms. T. A. Head (PE&RAS File)

Mr. B. H. Clark Mr. R. D. Martin Mr. W. F. Conway Mr. T. C. Morton Mr. G. W. Davis Mr. J. H. O'eill, Jr.

Mr. W. J. Dorman (BNP) Mr. J. S. Scarola Mr. R. S. Edwards Mr. J. M. Taylor Mr. R. J. Field Nuclear Records Mr. K. N. Harris Harris Licensing File Files: H-X-0511 H-X-0642

~>)'- Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 1 of 18 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION REGARDING THE ALTERNATIVEPLAN FOR SPENT FUEL POOL COOLING AND CLEANUP SYSTEM PIPING Re uested Information Item 1:

Explain how the Metorex X-Met 880 Alloy Analyzer discriminates between the different standards that you used in your analysis described in Enclosure 3, "Metallurgy Unit Report for Spent Fuel Pool Weld Metal Composition analysis," of your April 30, 1999, RAI response.

What are the chemical element ranges associated with the different standards that you used?

What determines a match on a particular standard? What chemical elements are not included in the "Match" determination and how are these elements reconciled?

I Res onse1:

The primary objective of the field alloy analysis was to confirm with reasonable assurance that the as-deposited weld material for the spent fuel pool piping field welds is an austenitic stainless steel material compatible with Type 304 stainless steel piping material. The chemical composition of the stainless steel filler materials are specified in ASME Section II, Part C, SFA-5.4/ 5.9. The elements controlled under this specification for stainless steel filler materials are:

carbon, chromium, nickel, molybdenum, columbium plus tantalum, manganese, silicon, phosphorus, sulfur, nitrogen, and copper.

The Alloy Analyzer was used in a comparison / identification mode. In the comparison /

identification mode, the unknown is compared to reference materials which are input by a specific measurement technique and stored in a memory location of the instrument. This method of analysis was selected to provide reasonable assurance that the chemical compositions of analyzed field welds are consistent with an austenitic stainless steel having a chromium content in the range of 18 to 24 weight percent and a nickel content in the range of 8 to 14 weight percent.

Explain how the Metorex X-Met 880 AlloyAnalyzer discriminates bet>veen the different standards that you used in your analysis described in Enclosure 4, "Metallurgy Unit Report for Spent Fuel Pool Weld Metal Composition Analysis," of your April 30, 1999, RAI response.

i) l 4 Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 2 of 18 The Metorex X-Met 880 Alloy Analyzer utilizes a Cadmium-109 isotopic source to excite the analyzed material and measure the secondary radiation produced by the source excitation. This, instrument can detect elements that range between and include chromium and molybdenum on the periodic chart of the elements. (The elements between and including terbium and uranium are also detected by this instrument with a cadmium source.)

The instrument was configured to detect six specific elements using the following pure element standards: (1) chromium, (2) manganese, (3) iron, (4) nickel, (5) copper, and (6) molybdenum.

Iron was selected because austenitic stainless steels are considered to be iron-based alloys; chromium, nickel, and molybdenum were selected because they are primary alloying elements; manganese was selected because it is a secondary alloying element; and copper was selected because it is a potential "tramp" (i.e., unwanted) element in this material that is detectable by this instrument. A backscatter standard was used to determine the background spectrum. The pure element standards and the backscatter standard were supplied with the instrument by the manufacturer. A series of comparison standards were loaded into the instrument for this analysis. These standards included: (1) Type 304 stainless steel, (2) Type 309 stainless steel, (3)

Type 310 stainless steel, (4) Type 316 stainless steel, and (5) NIST SRM 1154a. These four secondary standards and one National Institute of Standards and Technology (NIST) Standard Reference Material (SRM) were used because: (1) the instrument was used in a comparison mode, and (2) none of the SRMs available from NIST have compositions consistent with either Type 304, Type 308, or Type 309 stainless steels. NIST SRM 1155 (Type 316 stainless steel) and NIST SRM C1287 (Type 310 stainless steel - modified) were used also, as independent reference checks of the instrument during the field analysis.

In the comparison / identification mode, the unknown is compared to reference materials which are input by a specific measurement technique and stored in a memory location of the instrument.

The alloy analyzer has a multi-channel analyzer (MCA) having 256 micro channels. These micro channels represent a specific X-ray energy range (e.g., Channel 1 - 1 to 2 eV, Channel 2 - 2 to 3 eV, etc.). Each element has an average value for its excitation X-ray energy and, in practice, the actual response has a Gaussian distribution. Each pure element has a range, or window, consisting of several micro channels based on the full width at half maximum value of the Gaussian distribution. Therefore, counts detected in an element window are due to a detectable and measurable concentration of this element. The pure element standards and the austenitic stainless steel standards have different compositions. The response of the instrument varies with the concentration of a given element in a standard. The counts obtained for a standard by this instrument are proportional to the elemental concentration(s). Each standard will have a unique pattern (or "fingerprint") of counts in the selected element windows based on its chemical composition. The instrument discriminates between standards and unknowns based on the similarity of the instrument response (or counts detected) to the element windows for the stored standards.

What are the chemical element ranges associated >vith the different standards that you useil?

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'A Document Control Desk

' Enclosure 1 to SERIAL: HNP-99-172

. Page 3 of 18 The chemical element ranges for the standards used are shown below in Table 1. The NIST SRM (1154a) that was used to set-up the Alloy Analyzer has a chemical composition that is not within the chemical composition range for any standard UNS stainless steel alloy. However, the

'ickel and chromium contents of the NIST 1154a standard are similar to the nickel content of the Type 309 comparison standard and the chromium content of the Type 304 comparison standard, respectively. The remaining detectable elements in these three comparison standards are comparable and cannot be used to accurately differentiate between the various unknowns.

TABLE 1 Chemical Element Ranges for Standards Used to Set-up the Metorex Alloy Analyzer Standard Com osition, Wei ht Percent Chromium Manganese Iron Nickel Copper Molybdenum T e 304 18.28 1.48 bal 8.13 0.19 0.17 T e 309 22.60 1.63 bal 13.81 T e 310 24.87 1.94 bal. 19.72 0.11 0.16 T e 316 16.74 bal 10.07 0.11 2.06 NIST 1154a 19.31 1.44 bal 13.08 0.44 0.068 Chemical Element Ran es for Standards Used to Check the Allo Anal zer NIST C1287 23.98 1.66 bal. 21.16 0.58 0.46 NIST 1155 18.45 1.63 bal. 12.18 0.169 2.38 The tolerances for the chemical element ranges for the secondary standards (nominal Type 304, Type 309, Type 310, and Type 316 stainless steels) are not known. These secondary standards were provided with mill test reports for their chemical compositions, but the precise accuracy of these standards is not known because they are not certified as traceable to primary reference standards. However, the applicable ASTM standards for these alloys permit a major alloying element range of between 1 and 2.5 weight percent (e.g., carbon content - 0.08 weight percent maximum; silicon content - 1.00 weight percent maximum; nickel content - 8.00 to,10.50 weight percent maximum; etc.) without the applicable product analysis tolerances that depend upon the specific element and its relative concentration.

What determines a ~natch on a particular standard?

During a test, the Alloy Analyzer detects, measures, and compares the counts obtained for the specified elements in the unknown to those for the standards that have been loaded into the instrument (the specified elements are those that were loaded as pure element standards during the instrument set-up). The X-ray energy detection range for each of the specified elements is pre-set in the instrument and is based on physical constants related to the energy difference between electron shells in atomic structures. The number of counts in each pure element range is measured and compared to the counts for these elements in the known comparison standards.

The difference in counts between the unknown and the comparison standards is measured. The instrument is configured with three thresholds (or limits) for the difference in counts between the

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~] \ V Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 4 of 18 closest standard and the unknown. The least amount of difference between a comparison standard and the unknown is indicated by "GOOD MATCH." Ifthere are differences between the unknown and standard that do not meet the "GOOD MATCH"criteria, but the unknown is similar to one or more standards, the alloy analyzer will indicate "POSSIBLE MATCH." Ifthe difference in counts is too large, the instrument will indicate "NO GOOD MATCH."

What chenncal elements are not inclruled in the Match" deternnnation and ho)v are these elements reconciled?

The primary objective of the field alloy analysis was to confirm with reasonable assurance that the as-deposited weld material was an austenitic stainless steel material compatible with the Type 304 stainless steel piping material. The chemical compositions of stainless steel filler materials are specified in ASME Section II, Part C, SPA-5.4/5.9. The elements controlled under this specification for stainless steel filler materials are: carbon, chromium, nickel, molybdenum, columbium plus tantalum, manganese, silicon, phosphorous, sulfur, nitrogen, and copper.

The alloy analyzer was set up to detect the primary alloying elements: chromium, nickel, and molybdenum. In addition, the alloy analyzer was also set up to detect the secondary alloying element manganese, the tramp element copper, and the alloy base iron. The remaining elements addressed in the specification, but not detected by the alloy analyzer, are: carbon, columbium plus tantalum, silicon, phosphorous, sulfur, and nitrogen. None of these elements are capable of being detected with the Metorex Alloy Analyzer using a Cadmium-109 source either'ue to their relative concentration or their X-ray excitation energy. These secondary alloying elements, while important to the weldability characteristics of the filler material, are not as important to the performance of the weld in service with regard to strength and corrosion resistance.

Samples of three spent fuel pool cooling piping field welds were obtained by plant personnel and submitted to an external commercial laboratory for chemical analysis. The elements that were not determined by field analysis and those that were used in the identification mode of the field welds were measured by this laboratory and are shown in Table 2. Laboratory analysis of this representative sample substantiates the results of the field analysis and provides additional assurance that the chemical compositions of spent fuel pool field welds are satisfactory.

) 1 1 Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 5 of 18 TABLE< 2 NSL Chemical Analysis Results Identification 2-SF-36-FW-450 2-SF-38-FW-451 2-SF-71-FW-329 Alloy Analyzer 304 SS Possible NIST 1154a NIST 1154a Results Possible Possible NSL Chemical Anal sis Results Carbon 0.13 0.10 0.064 Niobium < 0.05 < 0.05 < 0.05 Chromium 20.08 20.11 19.06 Co er 0.054 0.10 0.093 Man anese 1.46 1.39 0.79 Mol bdenum 0.12 0.10 0.085 Nickel 9.30 9.24 9.63 Phos horus 0.021 0.021 0.026 Sulfur 0.007 0.005 0.013 Silicon 0.37 0.39 0.25 Titanium 0.011 In summary, the alloy analyzer was set up to confirm with reasonable assurance that the as-deposited weld material for the spent fuel pool piping field welds is an austenitic stainless steel material compatible with the reported Type 304 stainless steel piping material and the chemical composition requirements specified in ASME Section II, Part C, SFA-5.4/5.9. The programmatic and procedural controls which existed at the time of construction, augmented by the testing and analysis effort described above, provide reasonable assurance that the weld material for the spent fuel pool piping field welds is the proper weld material and will perform satisfactorily in service.

Re uested Information Item 2:

Provide assurance that the ferrite numbers are acceptable for A-No. 8 weld wire (ND-2433) used in welds with missing weld wire documentation.

Res onse2:

Ferrite numbers have been measured for 18 of the 19 accessible field welds remaining in the scope of the Alternative Plan (one field weld is located underneath a grating which could not be removed at the time the measurements were taken). The results of this work show mean ferrite numbers ranging from approximately 4 to 9 FN. SFA 5.9, Section A4.12 states that the ferrite potential for 308, 308L, and 347 is approximately 10 FN, but notes that the ferrite content may vary by +/- 7 FN or more around these midpoints and still be within the limits of the chemical

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Document Control Desk to SERIAL: HNP-99-172 Page 6 of 18 specification. Furthermore, Section A4.13 also states that the ferrite potential of a filler metal is usually modified downward in the deposit due to changes in the chemical composition caused by the welding process and technique used.

Ferrite is know to be beneficial in reducing the tendency for cracking or fissuring in weld metals; however, it is not critical, particularly under the mild service conditions associated with the spent fuel pool cooling system. Assurance that the ferrite numbers are acceptable is demonstrated by the following: (1) the measured ferrite numbers are reasonably consistent with those expected for the type of filler. material used, (2) all of the exposed field welds in the scope of the Alternative Plan have successfully completed a liquid penetrant examination which noted no evidence of cracks or fissures, (3) a strict materials control program governed issuance and control of weld materials, and (4) there is no evidence that incorrect or uncontrolled filler material might have been used.

Re uested Information Item 3:

Explain the chemical analysis in the Table associated with PQR 6(c), dated 11/15/84, page 2 of 2, laboratory test No. 9-2-149 described in Enclosure 6, "Lab Test Reports," of your April 30, 1999, RAI response. What row(s) are associated with the base material, weld, and standard(s)? What criteria was used to determine acceptability?

Res onse3:

Welding Procedure Specification (WPS) 8B2, Revision 16 is supported by four Procedure Qualification Records (PQRs). The original procedure qualification test, as documented on PQR 6, was performed in 1976. The procedure qualification test coupon for this test was prepared from 10 inch schedule 40 pipe, which has a wall thickness of 0.365 inches. This test coupon thickness supports a qualified base metal thickness range of 3/16 (0.1875) inches to 0.730 inches.

In 1981, an additional procedure qualification test, as documented in PQR 6(A), was performed to support the extended thickness range of 3/16 inches to 8 inches. This new qualified range was achieved by welding a 1.5 inch thick weld test coupon. In 1982, another procedure qualification test was performed, as documented in PQR 6(B), to expand the thickness range qualified to include a base material thickness as thin as 0.049 inches. This extended range was achieved by welding a 0.049 inch wall thickness test coupon. In 1984, the final procedure qualification test, as documented in PQR 6(C), was performed to extend the qualified thickness range to include materials as thin as 0.031 inches. This new thickness range was achieved by welding a weld test coupon with a thickness of 0.031 inches.

The portion of WPS 8B2, Revision 16 that was used to fabricate the fuel pool piping, based on base metal thickness range, is supported by PQR 6 and PQR 6(A). The fuel pool piping has a nominal wall thickness of 3/8 (0.375) inches, which is within the qualified base metal thickness range of 3/16 (0.1875) inches to 0.730 inches for PQR 6 and 3/16 (0.1875) inches to 8 inches for PQR 6(A).

Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 7 of 18 Relative to the chemical analysis in the Table associated with PQR 6(c), dated 11/15/84, page 2 of 2, laboratory test No. 9-2-149, referenced WPS 8B2 addresses welding of a SA240 TP 304 test coupon with a thickness of 0.031 inch. The documented mechanical test results reference two test specimens having a thickness of 0.031 inch (E&E Laboratory Test Number 9-2-149, specimen numbers 699 and 700). PQR 6(c) references an Arcos welding filler material, which according to the Certified Material Test Reports (CMTRs) attached to PQR 6(c) is Type 316 stainless steel filler material.

A definitive explanation for all of the entries on the data sheet in question, page 2 of 2 of the chemical analysis results, can not be provided due to insufficient documentation. However, based on the documentation supporting the procedure qualification test for PQR 6 (C),

Metallurgy Unit test records and anecdotal information, it appears that Harris Welding Engineering personnel requested the ERE Laboratories to perform mechanical testing and chemical analyses for a completed welding procedure qualification coupon performed using 0.031 inch thick Type 316 stainless steel base material. It is believed that the chemical analysis requested was to be performed on a sample of the material taken from the item that was to be welded in production and which provided the impetus to perform the additional weld procedure qualification. This is supported by the fact that chips of the supplied material were provided to the Analytical Chemistry Laboratory on November 12, 1984 (sampled on November 9, 1984) while the PQR is dated November 15, 1984. This indicates that the chemical analysis was performed prior to the welding of the procedure qualification test coupon and should not be considered a part of the procedure qualification test.

Re uested Information Item 4:

For the piping and welds examined internally, provide a discussion of the examination results.

What inspection criteria is used for evaluating the piping and welds for corrosion and fouling?

Describe the corrosion and fouling inspection procedure and inspection personnel qualification process. For the embedded welds not examined internally, describe what is preventing their

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examination. Discuss why the decision not to inspect all of the embedded welds will result in an acceptable level of quality and safety.

Res onse4:

An initial visual inspection of the embedded piping and welds was completed using a pneumatically-powered crawler carrying a high resolution camera. This crawler employed two sets of pneumatic cylinders which expanded and contracted in coordination with a single cylinder between them to produce an "inch worm" effect. Inspections of four of the eight embedded spent fuel pool cooling lines were performed using this crawler, including six embedded field welds.

Camera resolution was excellent and the visual inspection of the lines was thorough. This arrangement proved unsuitable, however, for longer lines having multiple elbows, and a decision was made to investigate other possible methods of inspecting the balance of embedded piping.

An arrangement was eventually selected which used flexible fiberglass rods to manually drive a camera on rollers through the pipe. A second inspection effort, only recently completed, used

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Document Control Desk Enclosure 1 to SERIAL: HNP-99-172

. Page 8 of 18 this crawler to successfully inspect all 9 of the remaining embedded field welds and associated piping.

The remainder of this response will focus on the initial inspection of four SFP cooling lines and six embedded welds. The results of the inspection of the remaining lines and nine embedded welds is still in the review process. Our preliminary evaluation is that the results of the second visual inspection are consistent with those of the first inspection and demonstrate that the piping and welds have not measurably degraded and are acceptable for their intended purpose.

The pneumatically-powered crawler provided a stable base from which to successfully complete a visual examination of the piping and welds which could be reached using this equipment. Each inspection was preceded by a resolution check wherein the camera was required to discern a 1.0 mil wire at the appropriate focal length, and the level of detail provided of the internal pipe surfaces was excellent. These inspections were conducted in accordance with Special Plant Procedure SPP-0312T, which provided specific acceptance criteria, as well as qualification requirements for the equipment and inspectors. The inspection included welds on four of the eight embedded cooling lines connected to Spent Fuel Pools C Ec D. All of the lines inspected were 12 inch, schedule 40 stainless steel (304) piping.

The initial inspection included the following field welds:

Field Weld Number Pi in Function 2-SF-8-FW-65 C SFP Cooling Supply 2-SF-8-FW-66 C SFP Cooling Supply 2-SF-143-FW-512 D SFP Cooling Supply 2-SF-144-FW-515 D SFP Cooling Supply 2-SF-144-FW-516 D SFP Cooling Supply 2-SF-159-FW-408 D SFP Cooling Supply In accordance with the acceptance criteria in Special Plant Procedure SPP-0312T, welds which can be accepted without further evaluation must be completely free of the following defects:

- no Cracks

- no Lack of Fusion

- no Lack of Penetration

- no Oxidation

- no Undercut greater than 1/32"

- no Reinforcement ("Push Through" ) greater than 1/16"

- no Concavity (Suck Back" ) greater than 1/32"

- no Porosity greater than 1/16"

- no Inclusions

Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 9 of 18 'age In addition, any indications not included in the above list of weld attributes but potentially pertinent to the condition of the piping and welds were required by the inspection procedure to be reviewed and formally evaluated by Harris Nuclear Plant Engineering staff. Such indications would include arc strikes, foreign material, evidence of mishandling, pipe mismatch, pitting, and evidence of corrosion.

The inspection procedure requires that personnel performing visual examinations be COL Visual Weld Examiners, certified in accordance with the Corporate NDE Manual. In addition, they are required to have successfully completed the CP&L training course on remote camera equipment and/or have demonstrated their capability to utilize the equipment to the satisfaction of the NDE VT Level III. Vendor personnel operating the closed circuit television system were not required to be certified visual weld examiners, but were required to be familiar with their equipment and proficient in its use.

Generally, the inspection results were good. It is noted that the welds in question were not subject to volumetric examination, and were sufficiently far from the open end of the pipe at the time of welding that an internal visual examination would not have been performed at the time of welding. Relative to the inspection criteria pertaining to weld attributes provided above, five of the six field welds were accepted based on the qualified examiner's review of the camera inspection video. A single weld, 2-SF-144-FW-516, was identified as having areas where portions of a consumable insert could be discerned. This weld, which exists in the horizontal O piping on the supply line to SFP D, had several locations where a consumable insert had been utilized but was not fully consumed. Generally, these locations were limited to several very small areas where a small portion of the insert could be discerned, but included one area about 1.5 inches long where a continuous portion of the insert could be seen.

The presence of a small amount of unconsumed insert is not considered to be an indication of an unqualified welder, inadequate procedures, or inappropriate materials. The small amount of unconsumed insert is a relatively insignificant imperfection which is not unusual on field welds such as 2-SF-144-FW-516, which was only subject to surface examination and does not lend itself to internal visual examination. ASME Section III, Subsection ND design rules recognize the potential for imperfections of this nature in welds not subject to volumetric examination, and require that a reduction in joint efficiency be assumed for butt welds which are subject to surface examination only (ref. ND-3552.2).

The root pass associated with the indication of unconsumed insert is backed up by multiple weld passes, any one of which would be adequate to establish a leak tight pressure boundary under these conditions. Hydrostatic test records show that field weld 2-SF-FW-144-516 successfully completed hydrostatic testing at 32 psi during construction prior to the line being embedded, and that this test was witnessed by both QC and the ANI. Procedures and processes at the time required that both these field welds were subject to multiple inspections and documentation reviews during construction. Given this, and considering that this weld was subject to multiple inspections at the time of construction, it is highly unlikely that the indications noted on field weld 2-SF-144-FW-516 extend into the root pass, let alone the multiple passes that followed it.

Document Control-Desk to SERIAL: HNP-99-172 Page 10 of 18 Since field weld 2-SF-144-FW-516 is on a line which connects directly to atmospheric spent fuel pools, hydraulic pressure at the welds is limited to static head and a small amount of friction losses. (The effect of velocity head would be sufficiently small as to be negligible, but would actually tend to reduce the effective pressure.) At the location of field weld 2-SF-144-FW-516, static head due to the elevation difference is approximately 286 - 277.5 = 8.5 feet. Piping friction losses per 100 ft for 12 inch steel piping is only about 3 feet at 4000 gpm, so even considering the effect of elbows in the line, the 55 foot length of piping between this field weld and SFP C would only contribute another few feet for a total head of about 10 feet (i.e., less than 5 psi).

Operation of the SFP cooling and cleanup system for the C & D pools will be at a relatively low temperature and very low pressure. Accordingly, the minimum wall thickness needed to retain this pressure over a localized area of reduced wall is only a very small percentage of the 0.375 inch wall thickness in this piping. The piping in the vicinity of field weld 2-SF-FW-516 is completely embedded in concrete, located approximately at the center of a six foot thick, seismically-designed wall. As such, this piping is not subject to externally induced movement or stresses. Since the SFP cooling and cleanup system operates at a relatively low temperature with little variation, thermally induced stresses and thermal cycling are not of appreciable concern.

Given the lack of externally induced stresses or thermal cycling, the small pieces of unconsumed insert will not initiate a crack or otherwise propagate a piping failure.

Based on all of the above considerations, the indications of an unconsumed insert identified on field weld 2-SF-144-FW-516 are acceptable, and no rework or repair to the weld is required.

Videotapes of the first six embedded field welds and associated piping to be visually inspected have been reviewed by CPS'ngineering and metallurgical personnel. Aside from localized occurrences of loosely adhering surface film (principally boron deposits from boric acid added to the water), the videotape provides clear evidence that the piping was free from fouling or foreign materials. Where necessary, deposits were removed with pressurized water before the visual inspection. It is the consensus of the reviewers that the condition of the piping and welds is very good. Several inconsequential stains and small pits were noted, indicating that a small amount of minor corrosion may have occurred at some time in the past. Videotapes of all 15 embedded field welds and associated piping have been forwarded to corrosion experts both within CP8cL and in the industry.

Re nested Information Item 5:

What are the chemical analyses for steel welds 2-CC-3-FW-207, 2-CC-3-FW-208, and 2-CC FW-209?

Res onse5:

Chemical analyses for the carbon steel chips have been completed and are provided as Enclosure 2 to this RAI response. The results of these analyses substantiate that the filler material used for these welds is generally consistent with chemical composition requirements found in SFA 5.1 for ER70S-6 and SFA 5.18 for E7018.

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'ocument Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page ll of 18 Re uested Information Item 6:

Describe the paper trail that identifies a specific weld material to a specific weld on the isometric drawings, i.e., show that the weld material being verified with the Metorex X-Met 880 was specified for that location. Identify missing documentation that breaks the paper trial, if any.

R~es ense 6:

The weld metal to be used on a given weld was prescribed by the Weld Procedure Specification.

The Weld Data Report (WDR) documented the Weld Procedure Specification to be used, as well as the AWS Classification of filler material. For the field welds for which WDRs are no longer available, it is not possible to directly document the Weld Procedure Specification and filler metal that was used. However, since the vendor data sheets are available on the pipe spools, a review has been done of the Weld Procedure Specifications available at that time and which would have been applicable for this type piping, material, and end prep. These Weld Procedure Specifications were provided to the NRC as Enclosure 6 to HNP-99-069, dated April 30, 1999, the HNP response to the March 24, 1999 NRC RAI on the Alternative Plan.

The pipe spools utilized in the HNP spent fuel pool cooling system are Type 304 stainless steel, a P-8 material. The Weld Procedure Specifications for P-8 to P-8 piping welds such as these in the spent fuel pool cooling system would have used filler metals conforming to SFA No. 5.4/ 5.9, including ER308, ER308L, ER316, ER316L and ER347. For Type 304 to Type 304 piping, ER308 would have typically been specified on the WDR. Given that some chemical changes in composition will be caused by the welding process and that blending of the base metal and filler metal would occur, the Metorex X-Met 880 testing is not intended to confirm the that chemical composition conforms to chemical composition requirements for each element, but rather to assure that weldments are sound by substantiating that the filler metal used was compatible with the piping material and generally consistent with composition requirements of the Weld Procedure Specification. Additional details on the use of the Alloy Analyzer to evaluate filler metal is provided in the HNP response to Requested Information Item 1 above.

Re uested Information Item 7:

Discuss the chemical analysis and any other analysis performed on the water in the fuel pool cooling and cleanup system (FPCCS) and component cooling water system (CCWS) for spent fuel pools (SFPs) C and D. Where did the water come from? Discuss any differences between the chemical analysis and the original water source. Provide the staff with a representative analysis of the water.

Res onse7:

A review of plant documentation substantiates that the embedded lines connected to SFPs C A D had water in them on two separate occasions during the construction process. Water samples were collected from seven of the eight lines associated with the embedded piping.

Analysis results of those water samples substantiate that the water in these lines originated from the spent

Document Control Desk to SERIAL: HNP-99-172 Page 12 of 18 fuel pools. Specifically, chloride and fluoride concentrations were very low, and generally consistent with specifications for spent fuel pool chemistry. Sulfate levels and conductivity, while not typically analyzed for spent fuel pool chemistry, were also very low and consistent with high purity water. The water samples also showed low levels of tritium, at a concentration similar to that of the spent fuel pools. Enclosure 3 to this RAI response provides a representative analysis of water samples taken from both the C and D SFP piping.

Initially, these lines were filled with water for hydrostatic testing prior to pouring concrete.

Potential sources of hydrotest water included potable water and lake water, although procedures did require that the piping be drained and vented subsequent to test completion. Since these lines could not be isolated from their respective fuel pool liners, they would have been filled again in support of pool liner leak testing. The procedure for liner leak testing required test water to have a chloride content of no more than 100 ppm, which effectively precluded the use of either potable water or lake water for this evolution. Furthermore, procedures required the pools to be drained after testing, then rinsed with distilled or demineralized water. Subsequent to liner leak testing, there was no reason to introduce water into the pools again until they were filled and put into service (1989 - 1990 time frame). Several of these lines were drained one additional time in 1995 - 1996, when drain valves were added to the exposed portions of several of the embedded lines. Since that time, these lines refilled with water from the spent fuel pools. The water samples that were collected and analyzed, as discussed above, were samples of water that leaked past "plumbers plugs" in the pool nozzles since this last evolution.

One of the eight lines has no drain line with an isolation valve for taking water samples, and was not represented in the initial set of water samples.

Re uested Information Item 8:

In Enclosure 8, "Hydrotest Records for Embedded Spent Fuel Pool Cooling Piping and Field Welds," of your April 30, 1999, RAI response, you provided signed hydrostatic test reports for 13 embedded welds. Starting with the signed hydrostatic test report, back track through procedures and program requirements to the point where the missing document(s) were verified as being complete. In other words, identify the specific procedural and program controls requiring verification of completion of the missing documentation (manufacturing/fabrication records, weld data records, updated isometric drawings, and inspections) starting backward from the hydrostatic test report.

Res onse8:

Construction procedure WP-115, "Pressure Testing of Pressure Piping (Nuclear Safety Related),"

governed the hydrostatic testing of the embedded lines connected to HNP SFPs C and D. This procedure specifically required, prior to hydrotesting, the Mechanical QA Specialist verify that:

1) all required piping documentation is complete, and

2) all required weld documentation is complete.

Document Control Desk to SERIAL: HNP-99-172 Page 13 of 18 Reference to piping and weld documentation is found in WP-102, "Installation of Piping."

Specific requirements found in this document include:

1) that each weld joint for Code piping receive a WDR, and that these WDRs receive a QA and ANI inspection.

2) that weld procedures utilized be qualified in accordance with MP-01, "Qualification of Weld Procedures."

3) that welders and welding operators be qualified in accordance with MP-02, "Procedure for Qualifying Welders and Weld Operators."

4) that welds be stamped in accordance with MP-05, "Stamping of Weldments."

5) that weld material be controlled in accordance with MP-03, "Welding Material Control."

Generally, items 2 - 5 above ensure that Code welds were performed to appropriate procedures in the plant'sSection IX weld program. Relative to item 1, WP-102 provided reference to CQC-19, "Weld Control" which again required that all Code welds received a WDR, and referenced procedure CQI-19.1, "Preparation Ec Submittal of Weld Data Report & Repair Weld Data Report," for detailed instructions on the use of WDRs. As prescribed by this procedure, the WDR included essentially all of the required attributes and documentation for welds within Code boundaries. Enclosure 4 provides a copy of CQI 19.1 at a revision level existing at or about the time most of the welds in question were made. Similarly, WP-102 contained requirements for layout and dimensional tolerances, as well as references to appropriate procedures for other piping installation processes, such as performance of cold pulls and torqueing of flanged connections. Therefore, in order to satisfy the prerequisites of procedure WP-115, the Mechanical QA Specialist would be required to verify that all the WDRs and RWDRs were complete and approved, dimensional and tolerance inspections had been completed, and all other piping installation processes had been completed and appropriately documented.

Re uested Information Item 9:

Identify the concrete pouring procedure that requires checking for the welder symbol and a successful hydrostatic test before pouring.

Res onse9:

Since embedding a line in concrete represented a point at which piping was no longer accessible for inspections, rework, etc., procedural controls were established to ensure that all required work activities had been completed and that documentation was in order prior to authorizing concrete placement. Procedure WP-05, "Concrete Placement", included a pre-placement requirement for a craft superintendent sign-off on the concrete placement report to signify completion of the craft's installation and superintendent inspection thereof. This procedure required that this sign-off be made by all craft superintendents, as a safeguard against omissions, whether or not they had material in a particular placement. Subsequently, procedure WP-05 required that the Construction Inspection Unit (QC) be notified when the installation was complete and ready for pre-placement inspection.

Document Control Desk to SERIAL: HNP-99-172 Page 14 of 18 Procedure TP-24, "Mechanical Pipe Installation Inspection" provided requirements for the Construction Inspection Unit relative to inspection of piping, and included separate sections on embedded piping inspection. This procedure specifically required the CI inspector to inspect the installation and documentation prior to concrete placement. The CI inspector was required to verify the specific installation attributes:

1) that piping installation was performed in accordance with design drawings and documents, notably including verification of pipe spool identification

2) that piping was free from physical damage, and had no missing parts, and

3) that all piping leak tests were complete and documented.

It can be seen that procedures associated with concrete placement did provide assurance that piping embedded in concrete was the correct piping and was correctly installed. Furthermore, since the hydro-test was generally the final milestone for completion of a pipe segment, verification that all piping leak tests were complete and documented provided assurance that all test and inspection requirements were met. Procedures WP-05 and TP-X do not specifically require a verification of the welder symbol. Rather, this assurance is provided by the review of weld documentation prior to hydro-testing, as well as the programmatic controls in CQC-19 and related procedures discussed above.

Re uested Information Item 10:

Describe how the liner leak tests support weld integrity for welds 2-SF-8-FW-65 and 2-SF FW-66 (Enclosure 3 of your response to NRCs RAI). For these two welds, back track through procedures and program requirements to the point where the missing documents were verified as being completed.

Res onse10:

Leak testing of the liner was accomplished under procedure TP-057, "Hydrostatic Testing of Fuel Pool Liners." This procedure provided specific steps to be completed prior to performance of the liner leak test. The procedure required that Engineering prepare the test package, including identification of all boundaries and all isolation points to be utilized. For the north spent fuel pool liner hydrostatic test, the documented test boundaries included the piping runs containing 2-SF-8-FW-65 and 2-SF-8-FW-66.

Subsequent to preparation of the test package, QC was required to complete the "Prerequisites" section of the test form. Similar to the discussion of piping hydro-test procedures provided in the response to Requested Information Item 8 above, these prerequisites included a line item for the QC Inspector to verify "all weld documentation complete." Although the test procedure was specifically concerned with inspection of the liners, this verification would have necessarily extended to the entire pressurized boundary to ensure that no external leakage occurred, that partially completed welds were not overstressed, etc.

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Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 15 of 18 Although hydrostatic test packages have not been located at this time for welds 2-SF-8-FW-65 and 2-SF-8-FW-66, plant documentation does support that this hydrostatic test was done. For example, QA Deficiency and Disposition Report (DDR) 794 was initiated to assess hydrostatic test requirements for the plate rings reinforcing the piping to pool nozzle connections. The resolution to this DDR acknowledged that the pipe spools adjacent to these welds had been subject to hydrostatic testing, even going so far as to include the dates of test performance. Four of the ten spools listed are included in the scope of the SFP C and D embedded piping, and two of these spools are in the line in which welds 2-SF-8-FW-65 and 2-SF-8-FW-66 are located. The other two spools referenced are on isometric drawing 2-SF-159, and are specifically included in a hydrostatic test package for which records have been located (provided previously to the NRC as Enclosure 7 to HNP-99-069, dated Apri130, 1999). Comparison of the dates listed on DDR 794 against those associated with piping on isometric drawing 2-SF-159 verify that the test dates on these documents are in agreement.

Therefore, even though hydrostatic test records specifically listing welds 2-SF-8-FW-65 and 2-SF-8-FW-66 as inspection items have not been located, it can be established with a high level of confidence that these welds were hydro-statically tested, and that documentation associated with these welds was reviewed and verified as being complete.

Re uested Information Item 11:

Describe precautions that were taken to protect system components (e.g., pumps, valves, heat exchangers, piping) from deleterious environmental effects during layup. Describe the layed up condition of the partially completed piping system and how this was determined. How would these layup conditions be different if it was known that SFPs C and D would be put in service later?

Res onse11:

The location of system components (e.g., pumps, valves, heat exchangers, piping), the area of the Fuel Handling Building, is fully enclosed and serviced by a safety related 236'levation HVAC system. This area is also the location of the operating Unit 1 spent fuel pool cooling pumps and heat exchangers, and is completely suitable for the long term storage of piping and equipment. It was anticipated that at some time it would be necessary to place C and D pools into service, and consideration was given to specific requirements for equipment protection. The spent fuel pool cooling pump motors were removed and placed in controlled storage conditions with heaters energized and shafts periodically rotated. The spent fuel pool heat exchangers were capped to preclude introduction of foreign material, and provided with a nitrogen blanket on the shell (CCW) side to prevent moisture and other contaminants from inducing corrosion. Spent Fuel Pool Cooling piping not connected to the spent fuel pools, which had never been wetted and was not connected to any active water systems, also received Foreign Material Exclusion (FME) type covers. Notably, the spent fuel pool cooling pumps and strainers were protected by FME covers on adjacent piping.

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Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 16 of 18 Through conversations with cognizant personnel, it is known that when it became necessary to fillthe C and D spent fuel pools, the exposed ends of the connected spent fuel pool piping were fitted with leak tight covers and flooded as well. At some point, "plumber's plugs" were fitted in the C and D spent fuel pool cooling nozzles, although it is not clear whether these plugs were installed before or after the lines were flooded by the spent fuel pools. The primary purpose of these plugs was not for equipment protection but instead for ALARAconsiderations, i.e., to preclude collection of radioactive material in the piping.

Re uested Information Item 12:

Why was visual inspection rather than ultrasonic inspection chosen to examine the integrity of the embedded welds?

Res onse12:

Examination requirements for the embedded spent fuel pool cooling piping at the time of construction consisted of a surface visual and liquid penetrant examination of the piping OD, consistent with design rules and NDE requirements in ASME Section III, Subsection ND.

Numerous programmatic and documentation assurances exist to confirm that these required inspections were indeed completed. In reviewing options for inspection of embedded piping and associated welds under the Alternative Plan, the objective was to implement an inspection program which: (1) provided yet another measure of assurance of construction quality, (2) provided a means to inspect as much of the overall scope as possible, (3) allowed for inspection of not only discrete areas of interest (ie., field welds), but also for qualitative assessment of overall piping condition, including corrosion and fouling, and (4) had a high. level of probability to produce meaningful results with existing, proven technology. These criteria are individually discussed as follows:

1) Provides additional measure of assurance of construction quality A detailed inspection of the interior of the piping with a high resolution camera provides a means to discern and assess numerous attributes pertaining to construction quality, including fit-up and alignment, adequacy of purge, and fusion of the root pass. These attributes, while readily examined with the use of a remote camera, do not lend themselves to detection and evaluation through ultrasonic examination.

2) Provides a means to inspect as much of the overall scope as possible Camera inspection provides a means to see as much of the overall inspection scope (piping interior surfaces) as possible, as well as focus on specific areas of interest. A number of vendors offer inspection services of piping using remote cameras and a variety of propulsion methods, providing the best probability of inspecting as much of the piping as possible. Using real time feedback, direct camera operators can move relatively quickly over long runs of piping which can be readily observed as clean and in good condition; however, considerable time is spent in adjusting focus, lighting and other parameters to provide a detailed examination of specific areas

Document Control Desk to SERIAL: HNP-99-172 Page 17 of 18 of interest. Although ultrasonic techniques are commonly used to detect wall thinning in steam piping, this process requires that the entire surface to be examined be mapped, with each grid location receiving an ultrasonic examination. Clearly, the lack of access in the embedded piping precludes the use of a similar technique to assess the overall condition of the embedded piping.

3) Allows for inspection of overall piping condition, but also macroscopic examination for fouling, corrosion, etc.

Camera inspection is the only viable means to identify and assess numerous attributes which pertain to the suitability of piping for service, including surface corrosion, fouling, foreign objects in the line, etc. Visual inspection with a high resolution camera can also detect visual evidence of corrosion (stains, discoloration) even when no loss of material or other degradation is obvious.

(4) Provide a high level of probability of producing meaningful results with existing, proven technology While not deemed appropriate to evaluate macroscopic examination of piping quality for the reasons discussed above, CPkL has investigated the feasibility of using ultrasonic examination to disposition discrete, localized indications. The obstacles associated with remotely performing ultrasonic examinations of these 12 inch embedded lines are considerable, and include:

Piping runs approaching 100 feet long Piping runs including 4 or more elbows Both horizontal and vertical runs Since pools are full, inspections must be done from the exposed piping end, meaning that all vertical runs are upward The weld joints themselves are irregular to the extent a direct beam method could not be used. In addition, these butt welds utilized consumable inserts with an end prep having a counterbore approximately '/4 inch from the weld joint. This configuration complicates the use of angle beam ultrasonic methods The piping surface must be clean and smooth, such that boron crystals or any other film or material which are in the area to be inspected must be removed.

A means must be devised to inject couplant in the area to be inspected The technique must provide a means to precisely locate and control the detector transducers, which would invariably require the use of a remote camera The device would need to be capable of propelling a camera, UT transducers, and all attendant cabling through long pipe sections with numerous elbows and risers to the location of interest, identify and focus on the indication to be examined, clean it as necessary, inject couplant on the area where the transducer will be placed, then precisely locate the transducer at that point, adjusting it as necessary to provide a good signal. Even then, since the back (outside) surface of the weld joints is irregular, it is not certain that the results will allow an accurate interpretation of the condition of the piping. In summary, while several vendors have expressed an interest in working on a cost and materials basis to provide the propulsion, robotics, and equipment

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Document Control Desk Enclosure 1 to SERIAL: HNP-99-172 Page 18 of 18 necessary to perform ultrasonic examination of the embedded piping, none have been identified with the proven experience necessary to provide repeatable, reliable results under similar conditions.

Re uested Information Item 13:

Describe the post modification testing to be performed to ensure that the system(s) will satisfy all design requirements. Include description of hydro-tests to verify the integrity of the system pressure boundaries, flushing to ensure unobstructed flow through the system components, and pre-operational functional testing under design flow/heat loads.

Res onse13:

Post modification testing will include the following:

1) System Hydrostatic testing conforming to Section IIIrequirements will be performed on the completed system. With the exception of embedded piping, components inside Code boundaries will be included in this test effort, including pumps, heat exchangers and strainers. In a previous HNP response to the NRC RAI on the Alternative Plan (ref. HNP-99-069, dated April 30, 1999), CP&L stated that Code Case N-240 would be used to exempt formal requirements for hydro-testing of the embedded piping connected to the atmospheric spent fuel pools. CP&L is continuing to investigate methods to provide additional assurance of the quality of embedded piping and field welds, including consideration of pressure testing. The final disposition of hydrostatic testing of embedded spent fuel pool piping will be provided to the NRC as part of the follow-up report on embedded piping and welds as discussed in the response to Requested Information Item 4 above.

2) A flush procedure will be developed which ensures that piping and components inside Code boundaries are free from fouling and debris which might affect system performance, reliability or spent fuel integrity.

3) Pre-operational testing will include a flow balance and verification which ensures that design flow requirements are met for the Spent Fuel Pool Cooling and Component Cooling Water systems, as well as those heat loads which rely on CCW (such as RHR) and heat sinks downstream of CCW (ESW, UHS). Given the lack of a heat load which would facilitate the performance of a meaningful heat duty test of the Spent Fuel Pool Cooling System, no such test will be performed. Moreover, at the 1.0 Mbtu / hr maximum heat load associated with this license amendment request, performance of such a test would not be viable even at the proposed licensed limit. Although the C and D spent fuel pool cooling heat exchangers were installed in the Fuel Handling Building nearly 20 years ago, they have never been placed into service and, from a design perspective, are still new. Moreover, these heat exchangers were layed up with a nitrogen blanket on the shell side, protecting it from moisture and corrosion.

A pre-service inspection of the tubesheets and tubes has been performed on these heat exchangers to ensure that no foreign material or corrosion exists which might obstruct flow or otherwise reduce performance.

KNCLOSVRK 2 to SERIAL: HNP-99-172 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION REGARDING THE ALTERNATIVEPLAN FOR SPENT FUEL POOLS C 8r D COOLING AND CLEANUP SYSTEM PIPING Carolina Power A Light Company Material Services Section Metallurgy Services Technical Report

Subject:

HNP - Material Identification of Chips from Carbon Steel Welds Associated with the Spent Fuel Pool Activation Project (1 page total)

CAROLINAPOWER 4 LIGHT COMPANY MATERIALSERVICES SECTION METALLURGYSERVICES TECHNICALREPORT To: Mr. Jeff Lane Project Number: 99-134 Date: Au ust25 1999 Investigators: Reviewed by:

Robert Jordan Dann Brinkle Distribution: Approved by.

File/Metallur Services Supervisor, Meta gy Services

SUBJECT:

HNP- Material Identification of Chips from Carbon Steel Welds Associated with the Spent Fuel Pool Activation Project.

On July 8, 1999 three samples of chips were received from HNP personnel for chemical analysis. The chips were removed from Welds 2CC-FW-207, 208, and 209 on ASME Section III, Class 3 Piping used on the Component Cooling Water (CCW) System. Metallurgy Services personnel were asked to perform chemical analysis on the three samples.

On July 27, 1999 the three samples of chips were sent to NSL Analytical Services, Inc., in Cleveland, Ohio for analysis. A report of the analyses was received from NSL on August 16, 1999. The results of the analysis for each sample are listed in the table below and a copy of the results from NSL is attached.

ELEMENT SAMPLE 2CC-FW- SAMPLE 2CC-FW- SAMPLE 2CC-FW-207 (WEIGHT 208 (WEIGHT 209 (WEIGHT PERCENT) PERCENT) PERCENT)

Carbon 0.13 0.11 0.11 Chromium 0.028 0.031 0.027 Co er 0.035 0.018 0.018 Man anese 1.29 1.20 1.15 Mol bdenum 0.014 0.004 0.003 Nickel 0.028 0.016 0.014 Phos horus 0.021 0.014 0.013 Sulfur 0.011 0.012 0.013 Silicon 0.29 0.29 0.41 Vanadium 0.018 0.026 0.026

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ENCLOSURE 3 to SERIAL: HNP-99-172 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION REGARDING THE ALTERNATIVEPLAN FOR SPENT FUEL POOLS C R D COOLING AND CLEANUP SYSTEM PIPING Chemistry Sample Data Sheets from HNP Procedure CRC-001 (2 sheets total).

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Title:

Preparation & Submittal of Weld Data Report, Repair Weld Data Report, Tank Fabrication Weld Record & Seismic I Weld Data Report Initial Issue Date: March 16, 1981 (25 pages total)

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. NLRRER: IHITIPL ISSUE DATE l QCI-19. 1 Q A RFCOROS March 16, 1981 01812 61981 APPROVED BYl RECNKNDED BYl I

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. ight-hand margin of the revised page(s). Manual holder is to re )ace affected a s only. This record is to be retained behind the title page of the instruction.

Rev. Description Signatures Date I

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QUALITY'ONTROL SECTION QCI-1 .1 PREPARATION 8( SUBMITTAL OF MELD DATA REPORT> REPAIR MELD DATA REPORT> TANK FABRICATION V 4.2.1 (cont.)

4.2.2 A repair WDR is not required for the following conditions')

Veld def'ects -which occur during the in-process welding and which can be removed and reworked within the Veld,twacedure,.Specification (MPS) specified;on the original WDR (this includes slag; porosity; burn-through in the root.pass or backing ring; or root weld defect in the pipe I.D. or O.D.).

....b J.,> Rework.,requirwd 4o,,correct,~-,process..Aefzc ts -found by NDE performed "for information".

c) Where complete removal of the weld joint is the repair method used (a'new WDR will be issued in this case).

4.3 Seismic I Weld Data Re ort (SWDR) 4.3. 1 Seismic I structural welding with the exception of stud welding shall-be-documentedmnm SVDR (QA-34).

4.3.2 Repairs to Seismic I structural welds will be documented on the SVDR when the'following con'dXtions exist:

a) A rejectable.defect.~Maund,by.wisual-~spec tion-or other NDE at a specified holdpoint or completed weld.

b) Damage metal.

to base material requiring deposition of filler ~

4.3.3 Entries on the SVDR are not required for the following conditions:

a) Weld defects which occur during the in-process welding and which can be removed and reworked within the Veld Procedure Specification (VPS) specified on the SWDR (this includes slag, porosity, burn-through in the root, pass or backing strip or root weld defect in the structural item).

b) Rework required to correct in-process defects found by NDE performed for "information only".

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PREPARATION 4 SUBMITTAL OF WELD DATA REPORT f REPAIR WELD DATA REPORT g TANK FABRICATIO 5.1 (cont. )

The white and yellow copies of the WDR, along with the work package, are forwarded to the Welding QA/QC Specialist.

The Welding QA/QC Specialist, or his designee, reviews the MDR for essential information and mandatory holdpoints and inserts additional holdpoints, if required. The ANI vill assign addi-tional holdpoints, if he desires, sign and date the MDR, concurs with the data'given, and return it to the Melding QA/QC if he Specialist. QA shall keep the yellow copy .of the.MDR and send the white copy along with the work package to the Mechanical Engineering Group for transmittal to the field. The areas of responsibility in filling out the MDR are-'outlined"below:

(Numbers correspond with Exhibit 1)

Title Data

1. Turnover No. No. assigned by Startup Group Meld Eng.

2. Veld Joint Zone,-* Income tric, "Field .Meld.dfo .,

Record No. Obtained from Isometric Weld Eng.

3. System System Name or designation Obtained from Isometric Weld Eng.

Category System Category (ASME Class 1,2,3, Seismic I) Obtained

'rom Isometric Veld .Fng.

5....Zng.. Dwg. No. Drawing,No. Obtained from Isometric Weld Cng.

6. Fill Meta1 Type of Filler Metal (E 7018, Type 309, 308, 316, etc.) Meld Eng.

7. Design Line Design Line No. Identification No. from Isometric/Drawing Weld Eng.

8. Base Metal ASME Spec. and Grade of base Spec. material being joined. Obtained from Isometric or Line Lists Weld Eng.

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CWR tEtIT CAROLiNA Pa~R R LiGHT C(NPANY CORPORATE QJALiTY ASSURANCE DEPARTf 1ENT ENGiNEERiNG R CONSTRUCTrON QUA( STY ASSURANCE/'QUALITY EEKR QNTROL SECTION QCI-19. 1 PREPARATION h SUBMITTAI OF WELD DATA REPORT, REPAIR MELD DATA REPORT, TANK TIME'ABRXCATIONWELD RFCORD 5 SEISMIC WELD DATA I 5.1 (cont.)

Title Data Res onsibilit

16. Xnse. vice Xnservice Insp. if required for Inspection,, the.field weld is assigned by Welding Engineering. Meld Eng.

17;. Welding Eng. Signature of Welding Engineer Verification (or'is designee) indicating

- Date - --- 'concurrence'-with ho1dpoin'ts. Meld Eng.

18. ANX'eview Signature of Authorized Nuclear Inspector (or his designee) in-dicatingwoncurrence with ho1'd-points. ANX

19. Release for Signature of Melding QA/QC QA and Date Specialist (or his designee) in-dicatcing concurrence~th 'holdpoints and releasing WDR to construction.

(Date Signed) ..-QA/QC Welding

20. Welder(s) Symbol(s) of Welder(s) assigned Symbol to perform weldiag. "(QC Inspector verifie s welde r qualification at this point) . ...., QA/QC Xnspector -.

21. Items QC Inspection holdpoints checked (/) that are required by Code, Specificat;ion, Pro-cedures, Drawings, or Isometric Meld Eng.

QC Xnspection holdpoints checked Melding QA/QC (u) that are desgnated by QA in Specialist, addition to holdpoints checked

(/) by Welding Engineer. (Hold-points that do not apply shall be marked N/A.)

ANI Inspection holdpoints checked (V') to be wit;nessed by ANI ANX 1'I

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QUALITY CONTROL MCTION QCI-19. 1 PREPARATION h SUBMITTAL OF WELD DATA REPORT, REPAIR WELD DATA REPORT, TANK 5.2.4 (cont.)
Title Data Responsibilit
19. Ht/Lot No. Heat No. of Filler Metal and/or Lot No. assigned to Filler Metal QA/QC Inspector
20. Welder' Symbol assigned to Welder Symbol entered at time of welding.
Root QA/QC Inspector
21. Welder Symbol assigned .to Welder
'ymbol entered at time of welding QA/QC Inspector Intermediate
22. Welder' Symbol assigned to Welder, Symbol entered at-time of-welding. QA/QC"-Inspector Final 23; Repair The instructions for repairing Instructions the weld as assigned by Welding
. Engineer *- Weld "Er.g.
24. Item Holdpoints Engineer checked (/)
"that are required by QA in add"'tion to holdpoints checked
..Ave.ky .Welding Engineer, Hold-points that do not apply shall be marked N/A. QA/QC Welding ANI holdpoints checked (/)
to be witnessed by ANI. 'Hold-points that do not, apply shall be marked N/A. ANI
25. QA/QC Signature of Welding QA/QC Specialist Specialist (or his designee) indicating final acceptance of weld repair. Date signed. Welding QA/QC Specialist
26. ANI Signature and date of ANI (Code Meld) indicating RWDR was reviewed and accepted. Date signed, ANI
'I!'M'Atter..;.... l'0tltl~i:,..::"":i'"'"'"":-:-'"-
COf% lJBf CAROLINA PmR 8 LIGHT Cd@Ate CORPORATE UuALrn ASSURANCE RPAR~NT ENG INEER I NG 8 .CONSTRUCT ION 9UAL I TY ASSURANCE/ %CHER BMlilQM GuALm CONTROL bECTION QCI-19. 1 PREPARATION h SUBMITTAL OF VELD DATA REPORT) REPAIR WELD DATA REPORT, TANK FABRICATION VELD RECORD A SEISMIC I 4 5.2,5 QA accepted signature signifies that the item has been repaired and accepted in accordance with the applicable MP specification and NDEP specification.
5.3 Seismic I WDR (SWDR) 5.3.1 Tne SWDR (QA-34 form) is initiated by the discipline engineer in the case of pipe hangerswnd-structural items.
It is initiated by the cr aft foreman for cable tray, conduit and HVAC supports. The appropriate individual fills out pertinent information and forwards the SVDR to the welding ergineer if holdpoints are required.
5.3.2 The white and yellow copies of the SMDR, along with the work .packa~ mre -forwarded..to,.the... WeldingQA/QC Special is t or his designee.
5.3.3 The Weldmg XUL/QC Specialzs't'r"his "de'si'gnome", reviews 'the SWDR for ess'ential information and mandatory holdpoints
"'and deserts additional&oldpoints~,wequired.
5.3.4 The Welding QA/QC Specialist, or his designee, will ini-tial and date the SWDR and "sen'd'"the'white opy to-the applicable Engineering discipline or craf t.
5.3.5 The areas of responsibility fpr filling out the SWDR ai below: (numbers corr espond with number ed blockse*'utlined on Exhibit 1) 5.3.5,1 Pi e Han rs & Structural A. Discipline Engineer (or his designee)
1. Completes blocks 1 through 6
2. Identifies joints involving 1-1/2" and thicker base material and assigns pre-heat holdpoints (and fitup holdpoints,
'f applicable).
3. Signs and dates: Retains pink copy and forwards white copy and yellow copy to Weld'ng Engineer.
B. Melding Engineer (or his designee)
1. Completes blocks 7, 8 and 9.
2. Identifies joint type and assigns man-datory holdpoints.
~~p4~@~+grryw~~Ppgp ~~~@~+. pgg~g~ggpyp I 0 I CAR IJBfT CARa.iNA PmR R LIGHT Cm WV CORPORATE UuAL.In ASSURANCE DEPAR~NT ENGINEERING 8 CONSTRUCTION 9UALITY ASSURANCE/ RHBH<
Aeu in CONTROL bECTION QCI-19. 1 PREPARATION 4 SUBMITTAL OF WELD DATA REPORT) REPAIR WELD DATA REPORT) TANK TITLE'ABRICATIONWELD
~
RECORD & SEISMIC I WEL 5.3.5.1 (cont. )
3. Identifies joints which require PWHT.
4. Sign and dates; forwards yellow and white copies to Welding QA/QC.
C. Welding QA/QC Specialist (or his designee)
1. Reviews entries made by Engineers against applicable drawings and specifications.
2. Designates additional holdpoints as needed.
3. Initials and dates; retains yellow copy and forwards white copy to discipline W',\ I'A engineer. 5 I
D. '--Discipline -Engineer
1. Forwards white, copy with work package to the craft foreman.
'. Craft Foreman 1., Completes weldout of joints not requiring preheat or fitup inspectior..
2, Notifies Welding QA/QC when ready for
-<<-~reheat-and/or - fitup -.inspection.
3. Notifies Welding QA/QC when ready for full penetration root pass holdpoints.
copy'"'""
4. Signs and dates Se'clan .II of white when all welds are complete.
F. Welding QA/QC Inspector
1. Completes items 1 through 3 in Section III.
2. Performs preheat and fitup inspection as designated. (Releases for weldout/root pass when acceptable.)
3. Performs root pass visual inspection of full penetration joints.
a. Performs specified ND">>, or
b. initiates NDE Request to the NDE subunit.
c. Releases for weldout when acceptable.
4. Performs final visual inspection of all joints and records welder(s) symbol(s).
5. Performs specified Final NDE or:
a. Initiates NDE Request to the NDE subunit.
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COf%0LLED tEI'Lt CAROLINA PmR R LIerr CeeANY CCePORATE QUA( IVV ASSURANCE DEPARmNT ENGINEERING 8 CONSTRUCTION QUALITY ASSURANCE/ HUHKR QUAL ITY CONTROL SECTION QCX-19. 1 SUBMITTAL OF WELD DATA REPORT, REPAXR WELD DATA REPORT, TANK TITLE'REPARATION 5 FABRICATION L 5.3.5.2 (cont.)
4, Signs and dates: Retains pink copy and forward white and yellow copies to the Welding Engineer.
C. Melding Engineer (or his designee)
1. ~ter data. in..blocks 7 ard 8 for full penetration welds and joints involving 1-1/2" thick base material. Other perti-nent welding information will be entered in block 9.
.. 2. aligns and,gates,;,,.Xurwards, qhi te and ~>>ellow copies to Melding QA/QC.
D. Welding QA/QC Specia'ist (or his designee)
~ -1. Review entries'made..by,.engineers;against applicable drawings and documents.
2. Designates additional holdpoints as needed.
3. Initials and dates; reta'ins 'ye1,low copy and forwards white copy to the craft foreman>>
E: Craft Foreman I
Notifies QA/QC when ready for preheat and/
or fitup holdpoints.
2. Notifies QA/QC when ready for full penetra-tion joint root pass holdpoints.
3 ~ Signs and datesSection II of white copy and yellow copy when all welds are completed.
F. Me'ding QA/QC Inspector 1 ~ Completes items 1 through 3 in Section III.
2. Performs preheat and fitup inspection as de ignated. (Releases for weldout/root pass when acceptable. )
3. Performs root pass visual inspection of full penetration joints.
a. Performs specified NDE, or
b. initiates NDE Request to the NDE subunit.
c. Releases for weldout when acceptable.
4. Performs final visual inspection of all joints and records welder(s) symbol(s).
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COf% LENT CAROLiNA PmR a LiGHT CW WV CORPORATE 9uALirt'SSURANCE DEPARmNT ENG i NEER ING R CONSTRUCT lON 9UAL1TY ASSURANCE/ HUEHZ, Geu i'ONTROL ECTiON QCI-19. 1 PREPARATION h SUBMITTAL OF WElD DATA REPORT, REPAIR WELD DATA REPORT, TANK FAB 5.3.5.2 (cont.)
5. Performs specified Final HDE or:
a. Initiates NDE Request to the NDE subunit.
b. Initiates request for vacuum box testing, if Monitors specified'.
PWHT in accordance with CQC-20, if specified.
7.. Acceptable welds>aving the same inspec-tion and NDE requirements may be tested collectively. Quantities as shown on applicable drawings, will be indicated (i.e. (8) fillet welds or (';) flare bevel welds).. Unacceptable~oints will be listed and identified separately (i.e.
5/16" fillet Pc. 5 to Pc. 8 top). Rein-spection and acceptance will be indicated by listing the joint again in the same
" section of.. the. QA-34 *for m.
5.4 Tank Fabrication Meld Pecord (TFWR) 5.4.1 The TFMR (QA-32 form) is initiated by the Melding Engineer (or his designee) who.wi11&M1.M, the tank design and identification data; joint identification, the material thickness, joint type, specified holdpoints and weld procedures for each weld 'jo'int."'*'The TFWR"is forwarded to Welding QA/QC.
5.4.2 The Welding QA/QC Specialist (or his designee) reviews the TFMR for essential req'uirements and mandatory hold-points; designates additional holdpoints, as needed; and submits it to the ANI (Code Class tanks only) for review and designation of his holdpoints.
Title Data Res onsibili'
1. Unit No. Assigned to Unit which tank belongs. Weld Eng.
2. Tank I.D. Obtained from tank drawing. Weld Eng.
Number
3. ASME Code ASME Code Class 1, 2 or 3. Meld Eng.
Class
4. Drawing Obtained from drawing. Meld Eng.
Number A,+ '~Q+gff ppfgf~$$'I~g>'~+@~ WA (p <<,M".. (4'4 (
'J<<e'<< "-' 'lltgQ 60f(mal%
Ceo.Ie PmR 8 LIGHT CeemV CCep0RAm QuALITY ASSueece RPAR~m ENG I NEER ING R (ANSTRUCT ION QNL ITY ASSURANCE/ ROBER QUALITY Cmm0L RCTIOW QCI-19. 1 PREPARATION & SUBMITTAL OF WELD DATA REPORT, REPAIR WELD DATA REPORT, TANK TI&Ee FABRICATION WELD RECORD 8( SEI MI 5.4.2 (cont.)
Title Data Res onsibilit 5 ~ Meld Engr. Signature of Weld Engr. '(or his designee) initiating the Tank Fabrication Meld Record and date. Weld Fng.
6 ~ Weld Number I.D. No. of weld from drawing. Weld Eng.
7. Material Ob'tained"frm drawing. Weld Eng.
Thickness
~ J
8. Joint Type Obtained from drawing. W"'d Eng.
9. Weld Proc. 'ss'i'gned'"by We1d'Engr. '""'Weld Eng'."
and NDE Requirements
10. Required Assigned by Weld Engr. Veld Eng.
Holdpoints
11. Veld Symbol -.From- assi gned welder/a.) .- Foreman
12. Material Heat From WMR. Foreman I
13. QA/QC Signature and date of QA/QC Inspector Inspector verifying holdpojnts. ...,., QA/QC. Zns paction ,..
14. ANI Signature and date of ANI verifying and/or adding holdpoints. ANI
15. QA/QC Signature and date of QA/QC Specialist Specialist or his designee after completion of TFMR. QA/QC Spec.
6.0 KYHIBITS Exhi.bit 1, Weld Data Report (VDR)
Exhibit 2, Repair Weld Data Repor t (Repair WDR)
Exhibit 3, Tank Fabrication Weld Record (TFvD)
Exhibit 4, Seismic I Veld Data Report (SWDR)
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SYSTEM CAT. DRAWING / ISOMETRIC FIELD WELD I.O.
(3) (4) (5) (6)
BASK METAL SPEC. 8 GRADE PIPE/COMPONENT SZE WELDING PROCEDURE 8 REV. NO.
(7') (9)
MATER (THICKNESS
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WELDING EhGlt(EER 8 DATE ANI REVIEW 8 DATE QA/QC REVIEW 8 DATE (13) (14) (15
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(15) SIZE ("7) HT NO.
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WELDER S SYMBOL- ROOT S SYMBOL -INTERMEDIATE WELDER S SYMBOL- FINAL (20) (21) (22)
RERQR INSTRUCTiONS: (23)
ITEM HOLD PT. ANI ACCEPTE D QA/QC ACCEPTED DATE A/QC ANI (24)
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ML18017A927

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