ML20097D457

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Proposed Tech Specs Re ISI & Testing Program
ML20097D457
Person / Time
Site: Millstone Dominion icon.png
Issue date: 02/12/1996
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20097D453 List:
References
NUDOCS 9602120396
Download: ML20097D457 (46)
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Text

t i Docket No.50-423 B15482 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Inservice Inspection and Testing Program Marked-up Pages I

9602120396 960205 PDR ADOCK 05000423 P PDR February 1996

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N W r 5 , 1990 APPLICABILITY / Nb 2 3 -[

LhMITINGCONDITIONFOROPERATION(Continued) 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been perfomed.

The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to pemit the l completion of the surveillance when allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting l Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. ,

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. )

Inservice inspection of ASF Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves l shall be performed in accordance with Section XI of the ASME Boiler 1 and Pressure Vessel Code and applicable Addenda as required by h i 10 CFR Part 50, Section 50.55a g), except where specific written 1 reiiet has been granted by h t Commission cursuant to 10 CFR / h ,r g fart 50,Section50.55a(g)(6)(1);] ~

b. Surveillance intervals specified in Section XI of the ASME Boiler i and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing insoettien and testina activities activities Weekly At least once per 7 days i Monthly At least once per 31 days l Quarterly or every 3 months At least once per 92 days 1 Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities; )
. 1 MILLSTONE - UNIT 3 3/4 0-2 AMENDMENTNO.54,f a J

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REACTOR COOLANT SYSTEM BASES (Continued) while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50' above primary temperature.

The Maximum Allowed PORY Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class I, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI cf the ASME Boiler and Prgtsure Vestel Code and applicabl Addenda as required by 10 CFR 50.554Ig) except where specific written relief Qas been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code < 80 Edition and Addeada through hHnter, except where specific written relief has been granted pursuant to 10 CFR -

50.55a(g)(6)(1).f 3/4.4.11 REACTOR C0OLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function. ,

The reactor vessel head vent path consists of two parallel flow paths with redundant isolation valves (3RCS*SV8095A, 3RCS*SV8096A and 3RCS*SV8095B, 3RCS*SV80968) in each flow path. The pressurizer steam space vent path )<

consists of two parallel paths with a power operated relief valve (PORV) and PORV block valve in series (3RCS*PCV455A, 3RCS*MV800A and 3RCS*PCV456, 3RCS*MV8000B). i

)

The valve redundancy of the Reactor Coolant System vent paths serves to ,

minimize the probability of inadvertent or irreversible actuation while l 1

ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980.

I MILLSTONE - UNIT 3 B 3/4 4-15 AmendmentNo.ppf, esas

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2 PLANT SY5TEMS BASES 3/4.7.6 FLOOD PROTECTION The limitation on flood protection ensures that the service water pump cubicle watertight doors will be closed before the water level reaches the critical elevation of 14.5 feet Mean Sea Level. Elevation 14.5 feet MSL is the level at which external flood waters could enter the service water pump cubicle.

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION 3YSTEM l

BACKGR0Ulm The control room emergency ventilation system provides a protected environment j from which operators can control the unit following an uncontrolled release of radioactivity. Additionally, the s3 stem provides temperature control for the i control room during normal and post-accident operations.

The control room emergency ventilation system is comprised of the control room emergency air filtration system and a temperature control system.

The control room emergency air filtration system consists of two redundant systems that recirculate and filter the control room air. Each control room emergency air filtration system consists of a moisture separator, electric heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system.

Normal Operation A portion of the control room emergency ventilation system is required to operate during normal operations to ensure the temperature of the control room '

is maintained at or below 95'F.

Post Accident Operation 1 The control room emergency ventilation system is required to operate during i post-accident operations to ensure the temperature of the control room is  !

maintained and to ensure the control room will remain habitable during and following accident conditions.

MILLSTONE - UNIT 3 B 3/4 7-6 AmendmentNo.JJ7,l one

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. h;;;.ber 22,-1995 STSGR 3-21-95

. PLANT SYSTEMS BASES 1 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

BACKGROUfB (Continued)

The following sequence of events occurs upon receipt of a control building isolation (CBI) signal or a signal indicating high radiation in the air supply duct to the control room envelope. l

1. The control room boundary is isolated to prevent outside air from entering the control room to prevent the operators from being exposed to the radiological conditions that may exist outside the control room.

The analysis for a loss of coolant accident assumes that the highest i releases occur in the first hour after a loss of coolant accident.

2. After 60 seconds, the control room envelope pressurizes to 1/8 inch water gauge by the control room emergency pressurization system. This action provides a continuous purge of the control room envelope and prevents inleakage from the outside environment. Technical l Specification 3/4.7.8 provides the requirements for the control room '

envelope pressurization system.

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3. Control room pressurization continues for the first hour.
4. After one hour, the control room emergency ventilation system will be placed in service in either the 100% recirculation mode (isolated from ,

the outside environment) or filtered pressurization mode (outside air is l diverted through the filters to the control room envelope to maintain a positive pressure). The mode of service for the filtration will be based on the radiological conditions that exist outside the control room. To run the control room emergency air filtration system in the filtered pressurization mode, the air supply line must be manually opened.

APPLICABLE SAFETY ANALYSIS The OPERABILITY of the Control Room Emergency Ventilation System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent for the duration of the accident. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A,10 CFR Part 50.

NILLSTONE - UNIT 3 B 3/4 7-7 Amendment No.

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. Neve h 22, 1005 1 1 PT50R 34195  !

PLANT SYSTEMS l

BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

LIMITING C0lWITION FOR OPERATION Two independent control room emergency air filtration systems are required to be operable to ensure that at least one is available in the event the other system is disabled.

l A control room emergency air flitration system is OPERABLE when the associated:

a. Fan is OPERABLE;
b. HEPA filters and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functiors; and
c. moisture separator, heater, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

The integrity of the control room boundary (i.e., walls, floors, ceilings, ductwork, and access doors) is covered by LIMITING CONDITION OF OPERATION (LCO) 3.7.8.

APPLICABILITY l

In MODES 1, 2, 3, 4, 5, and 6. 1 ACTIONS l Modes 1. 2. 3. and 4 With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. In this condition, the remaining control room emergency air filtration system is adequate to perform the control room protection function.

However, the overall reliability is reduced because a single failure in the OPERABLE train could result in a loss of the control room emergency air filtration system function. The 7-day completion time is based on the low probability of a DBA occurring during this time period, and the ability of the remaining train to provide the required capability.

If the inoperable train cannot be restored to an OPERABLE status within 7 days, the unit must be placed in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and within COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. These completion times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

MILLSTONE - IMIT 3 B 3/4 7-8 Amendment No.

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c. u r ::,1;;r PT5Gi 3-21-n PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

ACT10MS (Continued)

Modes-5 and 6

a. With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days, or to initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the recirculation mode. Initiating and maintaining operation of the OPERABLE train in the recirculation mode ensures: (i) operability of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected,
b. With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'a' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position.

SURVEILLANCE RE0VIREMENTS l

4.7.7.a The control room environment should be checked periodically to ensure that the control room temperature control system is functioning properly.

Verifying that the control room air temperature is less than or equal to 95'F at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient. It is not necessary to cycle the control room ventilation chillers. The control room is manned during operations covered by the technical specifications. Typically, temperature aberrations will be readily apparent.

4.7.7.b Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing the trains once every 31 days on a ,

STAGGERED TEST BASIS provides an adequate check of this system. This surveillance requirement verifies a system flow rate of 1,120 cfm i 20%. i Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized. These operations are sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters due to the humidity in the ambient air.

MILLSTONE - LNIIT 3 8 3/4 7-9 Amendment No.

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. heat s 22, 1995 PTSCR 3-21-95

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PLANT SYSTEQ l

BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.c The performance of the control room emergency filtration systems should be checked periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. The frequency is at least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.

ANSI N510-1980 will be used as a procedural guide for surveillance testing.

4.7.7.c.1 This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05%

in accordance with Regulatory Position C.S.a. C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate i of 1,120 cfm i 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 I referenced in the regulatory guide. l 4.7.7.c.2 This surveillance requires that a representative carbon sample be I obtained in accordance with Regulatory Position C.6.b of Regulatory Guide l 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the j representative carbon sample meets the criteria of Regulatory Position C.6.a )

of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide ,

penetration of less than 0.175%. The laboratory analysis is required to be l performed within 31 days after removal of the sample. ANSI N510-1980 is used 1 in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52. l 4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm i 20%, l during system operation when testing in accordance with ANSI N510-1980. l 4.7.7.d After 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the criteria of Regulatory position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl MILLSTONE - UNIT 3 5 3/4 7-10 Amendment No.

0436

., Sc1;.tr 22, -M9fr JISCP3-21-95--

. PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCEREQUIREMENTS(Continued) iodide penetration of less than 0.175%. The laboratory analysis is required to be perfomed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

The maximum surveillance interval is 900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, per Surveillance Requirement 4.0.2. The 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation requirement originates from Nuclear Regulatory Guide 1.52, Table 2, Note C, which states that " testing should be performed (1) initially, (2) at least once per 18 months thereafter for systems maintained in a standby status or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operations, and (3) following painting, fire, or chemical M ease in any ventilation zone communicating with the system." This testing ensures that the charcoal absorbency capacity has not degraded below acceptable limits as well as providing trending data. The 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> figure is an arbitrary number which is equivalent to a 30 day period. This criteria is directed to filter systems that are normally in operation and also provide emergency air cleaning functions in the event of a Design Basis Accident. The applicable filter units are not normally in operation and sample canisters are typically removed due to the 18 month crf,teria.

4.7.7.e.1 This surveillance verifies that the pressure drop across the combined HEPA filters and charcoal adsorbers banks at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm i 20%. The frequency is at least once per 18 months. l 4.7.7.e.2 This surveillance verifies 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 230 cfm relative to adjacent areas during system operation. The frequency is at least once per 18 months.

The intent of this surveillance is to verify the ability of the control room emergency air filtration system to maintain a positive pressure while running in the filtered pressurization mode. This capability is independent from the requirements regarding the control room pressurization system contained in Technical Specification 3/4.7.8.

i MILLSTONE - UNIT 3 8 3/4 7-11 Amendment No.

o 3-21-95 PLANT SYSTENS BASES 3/4.7.11 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Naterial sources will not exceed allowable intake values.

Sealed sources are classified into three grou)s according to their use, with Surveillance Requirements commensurate with tie probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.14 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of 12.2*F.

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i NILLSTONE - UNIT 3 5 3/4 7-21 Amendment Nos. pp, pf, Jpp, JJp,

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Docket No. 50-423 B15482 1

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i Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Inservice Inspection and Testing Program Retyped Pages i i

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i February 1996

APPLICABILITY o LINITINGC00mITIONFOROPERATION(Continued) 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.

The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This privision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME l 3

Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a;
b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing insoection and testina activities activities Weekly At least once per 7 days Monthly At least once per 31 days  !

Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities; MILLSTONE - UNIT 3 3/40-1 Amendment No. JJ 77, 0438 1

REACTOR C0OLANT SYSTEN

- BASES while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP is secondary temperature is more than 50' above primary temperature.  :

't The Maximum Allow PORV Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a. l ,

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 80 Edition and Addenda through Winter.

3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function.

The reactor vessel head vent path consists of two parallel flow paths with redundant isolation valves (3RCS*SV8095A, 3RCS*SV8096A and 3RCS*SV8095B, 3RCS*SV809658) in each flow path. The pressurizer steam space vent path consists of two parallel paths with a power operated relief valve (PORV) and PORY block valve in series (3RCS*PCV455A, 3RCS*MV800A and 3RCS*PCV456, 3RCS*MV80008).

The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while  ;

ensuring that a single failure of a vent valve, power supply, or control I system does not prevent isolation of the vent path. l The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980.

MILLSTONE - UNIT 3 8 3/4 4-15 Amendment No. pp, 57, 0439

i PLANT SYSTEMS j

. BASES 3/4.7.6 FLOOD PROTECTION The limitation on flood protection ensures that the service water pump cubicle watertight doors will be closed before the water level reaches the critical elevation of 14.5 feet Mean Sea Level. Elevation 14.5 feet MSL is the level at which external flood waters could enter the service water pump cubicle.

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUW The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity. Additionally, the system provides temperature control for the control room during normal and post-accident operations.

The control room emergency ventilation system is comprised of the control room emergency air filtration system and a temperature control system.

The control room emergency air filtration system consists of two redundant j systems that recirculate and filter the control room air. Each control room

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emergency air filtration system consists of a moisture separator, electric heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, I valves or dampers, and instrumentation form part of the system.

Normal Ooeration A portion of the control room emergency ventilation system is required to operate during normal operations to ensure the temperature of the control room is maintained at or below 95'F.

Post Accident Ooeration The control room emergency ventilation system is required to operate during post-accident operations to ensure the temperature of the control room is maintained and to ensure the control room will remain habitable during and following accident conditions.

NILLSTONE - UNIT 3 B 3/4 7-6 AmendmentNo.JJJ,l 0436

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PLANT SYSTEMS j

. BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued) l BACKGROWS (Continued) 1 The following sequence of events occurs upon receipt of a control building isolation (CBI) signal or a signal indicating high radiation in the l air supply duct to the control room envelope.

l. The control room boundary is isolated to prevent outside air from entering the control room to prevent the operators from being exposed to the radiological conditions that may exist outside the control room. .

The analysis for a loss of coolant accident assumes that the highest '

releases occur in the first hour after a loss of coolant accident.

2. After 60 seconds, the control room envelope pressurizes to 1/8 inch water gauge by the control room emergency pressurization system. This action provides a continuous purge of the control room envelope and prevents inleakage from the outside environment. Technical Specification 3/4.7.8 provides the requirements for the control room envelope pressurization system.
3. Control room pressurization continues for the first hour.
4. After one hour, the control room emergency ventilation system will be placed in service in either the 100% recirculation mode (isolated from the outside environment) or filtered pressurization mode (outside air is diverted through the filters to the control room envelope to maintain a positive pressure). The mode of service for the filtration will be based on the radiological conditions that exist outside the control ,

room. To run the control room emergency air filtration system in the l filtered pressurization mode, the air supply line must be manually i opened.

APPLICABLE SAFETY ANALYSIS  !

The OPERABILITY of the Control Room Emergency Ventilation System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent for the duration of the accident. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A,10 CFR Part 50.

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NILLSTONE - UNIT 3 B 3/4 7-7 Amendment No.

0436

PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

LIMITING C0lWITION FOR OPERATION Two independent control room emergency air filtration systems are required to be operable to ensure that at least one is available in the event the other system is disabled.

A control room emergency air filtration system is OPERABLE when the associated:

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a. Fan is OPERABLE;
b. HEPA filters and charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions; and
c. moisture separator, heater, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

The integrity of the control room boundary (i.e., walls, floors, ceilings, ductwork, and access doors) is covered by LIMITING CONDITION OF 1 OPERATION (LCO) 3.7.8.

APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6.

4 ACTIONS Modes 1. 2. 3. and 4 With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. In this condition, the remaining control room emergency air filtration system is adequate to perform the control room protection function.

However, the overall reliability is reduced because a single failure in the OPERABLE train could result in a loss of the control room emergency air filtration system function. The 7-day completion time is based on the low probability of a DBA occurring during this time period, and the ability of the remaining train to provide the required capability.

If the inoperable train cannot be restored to an OPERABLE status within 7 days, the unit must be placed in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and within COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. These completion times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

MILLSTONE - UNIT 3 8 3/4 7-8 Amendment No.

0436

PUU(T SYSTEMS

. BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

ACTIONS (Continued)

Modes 5 and 6

a. With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days, or to initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the  ;

recirculation mode. Initiating and maintaining operation of the '

OPERABLE train in the recirculation mode ensures: (1) operability of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected.

b. With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'a' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving CORE ALTERATIONS or positive reactivity changes. j This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position. -

SURVEILLANCE REQUIREMENTS 4.7.7.a The control room environment should be checked periodically to ensure that the control room temperature control system is functioning properly.

Verifying that the control room air temperature is less than or equal to 95*F at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient. It is not necessary to cycle the control room ventilation chillers. The control room is manned during operations covered by the technical specifications. Typically, temperature aberrations will be readily apparent.

4.7.7.b Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing the trains once every 31 days on a STAGGERED TEST BASIS provides an adequate check of this system. This surveillance requirement verifies a system flow rate of 1,120 cfm i 20%.

Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized. These operations are sufficient to reduce i the buildup of moisture on the adsorbers and HEPA filters due to the humidity  !

in the ambient air. l l

MILLSTONE - UNIT 3 B 3/4 7-9 Amendment No.

0436

PLANT SYSTEMS

. BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.c The performance of the control room emergency filtration systems should be checked periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. The frequency is at least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.

l ANSI N510-1980 will be used as a procedural guide for surveillance testing.

1 4.7.7.c.1 i This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05%

in accordance with Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 1,120 cfm i 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in the regulatory guide.

4.7.7.c.2 This surveillance requires that a representative carbon sample be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the representative carbon sample meets the criteria of Regulatory Position C.6.a i of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide l penetration of less than 0.175%. The laboratory analysis is required to be l performed within 31 days after removal of the sample. ANSI N510-1980 is used  ;

in lieu of ANS1 N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm i 20%,

during system operation when testing in accordance with ANSI N510-1980.

4.7.7.d After 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the criteria of Regulatory position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl MILLSTONE - UNIT 3 B 3/4 7-10 Amendment No.

0436

4 2

PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIRENENTS (Continued) iodide penetration of less than 0.1/5%. The laboratory analysis is required to be performed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

The maximum surveillance interval is 900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, per Surveillance Requirement 4.0.2. The 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation requirement originates from Nuclear Regulatory Guide 1.52, Table 2, Note C, which states that " testing should be performed (1) initially, (2) at least once per 18 months thereafter for systems maintained in a standby status or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operations, and (3) following painting, fire, or chemical release in any ventilation zone communicating with the system." This testing ensures that the charcoal absorbency capacity has not degraded below acceptable limits as well as providing trending data. The 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> figure is an arbitrary number which is equivalent to a 30 day period. This criteria is directed to filter systems that are normally in operation and also provide emergency air cleaning functions in the event of a Design Basis Accident. The applicable filter units are not normally in operation and sample canisters are typically removed due'to the 18 month criteria.

4.7.7.e.1 This surveillance verifies that the pressure drop across the combined HEPA filters and charcoal adsorbers banks at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm i 20%. The frequency is at least once per 18 months.

4.7.7.e.2 This surveillance verifies 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 230 cfm relative to adjacent areas during system operation. The frequency is at least once per 18 months.

The intent of this surveillance is to verify the ability of the control room emergency air filtration system to maintain a positive pressure while running in the filtered pressurization mode. This capability is independent from the requirements regarding the control room pressurization system contained in Technical Specification 3/4.7.8.

NILLSTONE - UNIT 3 B 3/4 7-11 Amendment No.

0436

PLANT SYSTEMS

- BASES 3/4.7.7 CONTROL ROON EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued)

During the first hour, the control room pressurization system creates and maintains the positive pressure in the control room. This capability is verified by Surveillance Requirement 4.7.8.C, independent of Surveillance Requirement 4.7.7.e.2. Furthermore, ACTIONS A.2 and B.1 of Limiting Condition for Operation 3.7.8 requires that an OPERABLE control room emergency air filtration system be initiated and maintained in the recirculation mode following both control room envelope pressurization systems becoming inoperable (e.g., a breech in the control room envelope). Running the control room air filtration system in the recirculation mode with the control room emergency pressurization inoperable would prohibit the ability to create and maintain a positive pressure in the control room envelope, because no source of air would be available to pressurize the control room envelope. A CBI signal will automatically align an operating filtration system into the recirculation mode of operation due to the isolation of the air supply line to the filter.

After the first hour of an event with the potential for a radiological release, the control room emergency ventilation system will be placed in service in either the recirculation mode (isolated from the outside environment) or filtered pressurization mode (outside air is diverted through the filters to the control room envelope to maintain a positive pressure).

The mode of service for the control room emergency air filtration system will be based on the radiological conditions that exist outside the control room.

Alignment to the filtered pressurization mode requires manual operator action to open the air supply line.

4.7.7.e.3 )

. This surveillance verifies that the heaters can dissipate 9.4 i 1 kW when tested in accordance with ANSI N510-1980. The frequency is at least once per 18 months.

4.7.7.f l

Following the complete or partial replacement of a HEPA filter bank, the  ;

operability of the cleanup system should be confirmed. This is accomplished '

by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1,120 cfm i 20%.

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i MILLSTONE - UNIT 3 B3/47-12 Amendment No. )

0436 j

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PLANT SYSTEMS i

! . BASES l 3/4.7.7 CONTROL ROON EMERGENCY VENTILATION SYSTEM (Continued) ,

l
SURVEILLANCE REQUIRENENTS (Continued) )

j 4.7.7.o i Following the complete or partial replacement of a charcoal adsorber j bank, the operability of the cleanup system should be confirmed. This is accomplished by verifying that the cleanup system satisfied the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05%

in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant

, test gas while operating the system at a flow of 1,120 cfm i 20%.

3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM

{

BACKGROUND l The control room envelope pressurization system provides a protected

! environment from which operators can control the unit following an j uncontrolled release of radioactivity.

The control room envelope pressurization system consists of two banks of i air bottles with its associated piping, instrumentation, and controls. Each <

l bank is capable of providing the control room area with one-hour of air l following any event with the potential for radioactive releases.

l Normal Operation During normal operations, the control room envelope pressurization i system is required to be on standby. l Post Accident Operation

! The control room envelope pressurization system is required to operate i during post-accident operations to ensure the control room will remain habitable during and following accident conditions.  ;

The sequence of events which occurs upon receipt of a control building isolation (CBI) signal or a signal indicating high radiation in the air supply duct to the control room envelope is described in Bases Section 3/4.7.7.

APPLICABLE SAFETY ANALYSIS The OPERABILITY of the control room envelope pressurization system ensures that: (1) breathable air is supplied to the control room, instrumentation rack room, and computer room, and (2) a positive pressure is created and maintained within the control room envelope during control MILLSTONE - UNIT 3 8 3/4 7-13 Amendment No.

0436 l

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j PLANT SYSTENS BASES 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (Continued)

APPLICABLE SAFETY ANALYSIS (Continued)

I building isolation for the first hour following any event with the potential l for radioactive releases. Each system is capable of providing an adequate air supply to the control room for one hour following an initiation of a control i building isolation signal. After one hour, operation of the control room i emergency ventilation system would be initiated.

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(

LINITING C0lGITION FOR OPERATION t

' Two independent control room envelope pressurization systems are i required to be operable to ensure that at least one is available in the event 2

the other system is disabled.

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l A control room envelope pressurization system is OPERABLE when the j associated:

i a. air storage bottles are OPERABLE; and  ;

b. piping and valves are OPERABLE.

In addition, the integrity of the control room boundary (i.e., walls, floors, ,

ceilings, ductwork, and access doors) must be maintained.

APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6.

ACTIONS

a. With one control room envelope pressurization system inoperable, action ,,

must be taken either: (1) to restore the inoperable system to an "

OPERABLE status within 7 days, (2) to initiate and maintain operation of  !

an OPERABLE control room emergency air filtration system in the recirculation mode, or (3) to place the unit in HOT STANDBY within six hours and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

For ACTION 3.7.8.a.1, the remaining control room envelope pressurization system is adequate to perform the control room protection function. l However, the overall reliability is reduced because a single failure in  !

the OPERABLE train could result in a loss of the control room envelope  !

pressurization system. The 7-day completion time is based on the low j probability of a design basis accident occurring during this time period  !

and the ability of the remaining train to provide the required capability.

4 NILLSTONE - UNIT 3 8 3/4 7-14 Amendment No.

0436

PLANT SYSTEMS

. BASES 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (Continued)

ACTIONS (Continued)

For ACTION 3.7.8.a 2, initiating and maintaining operation of an OPERABLE train of the control room emergency air filtration system in the recirculation mode ensures that (1) any inleakage, as a result of i loss pressurization, will be filtered from the initiation of the event, i and (ii) active failures of that train will be readily detected. To meet the requirements of this action statement, the control room i emergency air filtration system could be manually placed in either the 1007, recirculation mode or the recirculation with makeup air mode. The recirculation with makeup air mode is used to refresh the control room air supply. While in the recirculation with makeup air mode, if a CBI l signal is received, the fresh air makeup would be automatically isolated and the filters aligned to the 1007, recirculation mode.

For ACTION 3.7.8.a.3, the completion times for the unit to be placed in l HOT STANDBY and COLD SHUTDOWN are reasonable. They are based on i operating experience, and they permit the unit to be placed in the  !

required conditions from full power conditions in an orderly manner and I without challenging unit systems.  ;

Stud tensioning may continue in MODE 6 and a MODE change to MODE 5 is permitted with a control room envelope pressurization system inoperabic (Reference 1).

b. With both control room envelope pressurization systems inoperable, I action must be initiated within one hour to restore one inoperable system to an OPERABLE status and either (1) initiate and maintain o>eration of an OPERABLE control room emergency air filtration system in tie recirculation mode, or (2) place the unit in HOT STANDBY within six hours and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

The rationale for ACTIONS 3.7.8.b.1 and 3.7.8.b.2 are the same as those for ACTIONS 3.7.8.a.2 and 3.7.8.a.3, respectively.

Solely due to inoperability of both trains of the control room envelope pressurization system, the conditions and required actions assigned with LCO 3.7.7 are not required to be entered.

ACTIONS a.2 and b.1 of Limiting Condition for Operation 3.7.8 require that an OPERABLE control room emergency filtration system be placed in the recirculation mode. Under normal plant conditions to meet this requirement, the system would be placed in service in the recirculation l

MILLSTONE - UNIT 3 8 3/4 7-15 Amendment No.

0436

PLANT SYSTEMS o BASES 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (Continued)

ACTIONS (Continued) with makeup air. This makeup air is used to refresh the control room envelope. In the event of a design basis accident (including control building isolation), with the filtration system operating in the recirculation with makeup air mode, the makeup air is automatically isolated and the filtration system goes into a 100% recirculation mode.

Although no positive pressure is maintained in this alignment, it ensures that unfilterable noble gases are not forced into the envelope.

The recirculation mode ensures that radiciodines introduced to the envelope are continuously filtered out. After one hour, the filters could be manually placed in the pressurization mode if radiological conditions permit.

SURVEILLANCE REQUIRENENTS 4.7.8.a This surveillance requires verification that the air bottles are properly pressurized. Verifying that the air bottles are pressurized to greater than or equal to 2200 psig will ensure that a control room envelope pressurization system will be capable of supplying the required flow rate.

The frequency of the surveillance is at least once per 7 days. It is based on engineering judgment and has been shown to be appropriate through operating experience.

4.7.8.b This surveillance requires verification of the correct position of each valve (manual, power operated, or automatic) in the control room envelope pressurization system flow path. It helps ensure that the control room envelope pressurization system is capable of performing its intended safety function by verifying that an appropriate flow path will exist. The surveillance applies to those valves that could be mispositioned. This surveillance does not apply to valves that have been locked, sealed, or secured in position, because these positions are verified prior to locking, sealing, or securing. '

The frequency of the surveillance is at least once per 31 days on a STAGGERED TEST BASIS. It is based on engineering judgment and has been shown to be appropriate through operating experience.

NILLSTONE - UNIT 3 B 3/4 7-16 Amendment No.

0436

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SURVEILLANCE REQUIREMENTS (Continued) 4.7.8.c The performance of the control room envelope pressurization system should be checked periodically. The frequency is at least once per 18 months and following any major alteration of the control room envelope pressure boundary.

A major alteration is a change to the control room envelope pressure boundary that: (1) results in a breach greater than analyzed for acceptable pressurization And requires nonroutine work evolutions to restore the boundary. A nonroutine work evolution is one which makes it difficult to determine As-Found and As-Left conditions. Examples of routine work evolution include: (1) opening and closing a door, and (2) repairing cable and pipe penetrations because the repairs are conducted in accordance with procedures and are verified via inspections. For these two examples, there is a high level of assurance that the boundary is restored to the As-Found condition.

This surveillance requires at least once per 18 months or following a major alteration of the control room envelope pressure boundary by:

  • Verifying, after a 60 second time delay following a Control Building Isolation Test signal, the control room envelope pressurizes to greater than or equal to 0.125 inch water gauge relative to outside atmosphere; and

Changes in conditions outside the control room envelope cause pressure I spikes which are reflected on the differential pressure indicator,  !

3HVC-PDI 113. I 1

Pressure spikes or fluctuations which result in the differential pressure momentarily dropped below the 0.125 inch water gauge acceptance criteria are acceptable providing the following conditions are met:

1. Differential pressure remains positive at all times.
2. Differential pressure is only transitorily below the acceptance i criteria.
3. Differential pressure returns to a value above the acceptance criteria.

MILLSTONE - UNIT 3 B 3/4 7-17 Amendment No.

0436

PLANT SYSTEMS

.- BASES 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued)

The control room envelope pressurization system design basis criteria is set at it 0.125 inch water gauge criteria to account for wind effects, thermal column effects, and barometric pressure changes. Pressurizing the control room envelope of 0.125 inch water gauge above the initial atmospheric pressure ensures it will remain at a positive pressure during subsequent changes in outside conditions over the next 60 minutes. Since the surveillance requirement is verified by actual reference to outside pressure, allowances are provided for differential pressure fluctuations caused by external forces.

The 0.125 inch water gauge acceptance criteria provides the margin for these fluctuations. This meets the requirements of Regulatory Guide 1.78 and NUREG-800, Section 6.4 and is consistent with the assumptions in the Control Room Operator DBA dose calculation.

4.7.8.c.1 This surveillance verifies that the control room envelope is isolated following a control building isolation (CBI) test signal.

4.7.8.c.2 This surveillance verifies that the control room envelope is pressurized to greater than or equal to 1/8 inch water gauge, relative to the outside atmosphere, within 60 seconds following receipt of a CBI test signal.

4.7.8.c.3 This surveillance verifies that the positive pressure developed in accordance with Surveillance Requirement 4.7.8.c.2 is maintained for greater than or equal to 60 minutes. This capability is independent from the requirements regarding the control room emergency filtration system contained in Technical Specification 3/4.7.7. Also, following the first hour, the control room emergency ventilation system is responsible for ensuring that the control room envelope remains habitable.

References:

(1) NRC Routine Inspection Report 50-423/87-33, dated February 10, 1988.

MILLSTONE - UNIT 3 B 3/4 7-18 Amendment No.

0436

i PLANT SYSTEMS BASES 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM The OPERABILITY of the Auxiliary Building Filter System ensures that radioactive materials leaking from the equipment within the charging pump, component cooling water pump and heat exchanger areas following a LOCA are filtered prior to reaching the environment. The charging pump / reactor plant component cooling water pump ventilation system must be operational to ensure operability of the auxiliary building filter system and the supplementary leak ,

l collection and release system. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to ,

reduce the buildup of moisture on the adsorbers and HEPA filters. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing.

3/4.7.10 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity ,

of the Reactor Coolant System and all other safety-related systems is main- l tained during and following a seismic or other event initiating dynamic loads, l For the purpose of declaring the affected system OPERABLE with the inoperable l snubber (s), an engineering evaluation may be performed, in accordance with ,

Section 50.59 of 10 CFR Part 50.  !

)

Snubbers are classified and grouped by design and manufacturer but not by size. Snubbers of the same manufacturer but having different internal  ;

mechanisms are classified as different types. For example, mechanical snubbers j utilizing the same design features of the 2-kip, 10-kip and 100-kip capacity -

manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydra 11c snubbers from  :

either manufacturer. 4 A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Plant Operations Review Committee. The determination shall be based upon the existing radiation levels and the expscted time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g.,

temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to each safety-related system during an earthquake or severe transient. Therefore, the required inspection interval varies inversely with the observed snubber failures on a given system and is determined by the number of ino)erable snubbers found during an inspection of each system.

In order to establisi the inspection frequency for each type of snubber on a l

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NILLSTONE - UNIT 3 B 3/4 7-19 Amendment No. 77, JJp, om

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PLANT SYSTEMS BASES 3/4.7.10 SNUBBERS (Continued) safety-related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncoverei, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing.

To provide assurance of snubber functional reliability, one of three functional testing methods is used with the stated acceptance criteria:

1. Functionally test 10% of a type of snubber with an additional 5%

tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or
3. Functionally test a representative sahiple size and determine sample acceptance or rejection using the stated equation.

Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by -

Acheson J. Duncan.

Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.

NILLSTONE - UNIT 3 B 3/4 7-20 Amendment Nos. JJ, 77, JJ7, one

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PLANT SYSTEMS r BASES 3/4.7.11 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.14 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of 2.2*F. ,

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MILLSTONE - UNIT 3 B 3/4 7-21 Amendment Mos. pp, pp, Jpp, JJp, one

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Docket No. 50-423

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B15482 t

Attachment 3 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications l Inservice Inspection and Testing Program Description of Proposed Changes l

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February 1996 l 1

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'. U.S. Nuclear Regulatory Commission B15482/ Attachment 3/Page 2 February 5, 1996 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Inservice Inspection and Testing Program Description of Proposed Changes The proposed changes would revise Technical Specification 4.0.5a,

" Surveillance Requirements for Inservice Inspection and Testing Program," and Bases Section 3/4.4.10, " Structural Integrity."

Specifically, the proposed changes relate to inservice inspection (ISI) and inservice testing (IST) requirements which are specified i in 10CFR50.55a, " Codes and Standards." The American Society of l Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (the Code) is incorporated by reference in the regulation as the  ;

requirements for ISI and IST. The proposed changes delete the phrase "(g), except where specific written relief has been granted ,

by the Commission pursuant to 10 CFR Part 50, Section !

50.55a(g) (6) (1) . " Similar changes are being proposed for the Bases Section. The proposed changes are consistent with the recommendations of NUREG-1482, " Guidelines for Inservice Testing at  !

Nuclear Power Plants." l l

The changes to Bases Sections 3/4.7.7 and 3/4.7.8 add design basis l information and provide clarification of system design and I operation to facilitate interpretation of the technical I specification action statements. These action statements cover plant operation when one or both trains of control room envelope pressurization are inoperable. These proposed changes only affect the Bases Section, therefore, they were reviewed in accordance with ,

the provisions of 10CFR50.59 and approved by the Millstone Unit l No. 3 Plant Operations Review Committee. The changes to Bases Sections 3/4.7.7 and 3/4.7.8 are being submitted to the NRC for information only.

Discussion NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power ,

Plants," specifically addresses the situation in which the i technical specifications are in conflict with the regulations of 10CFR50.,55a. As discussed in NUREG-1482, the NRC staff recognized that situations could arise which would put the licensee in a condition that is not in strict compliance with Technical Specification 4.0.5 requirements to comply with ASME Section XI "except where specific written relief has been granted." According to the NUREG, if Technical Specification 4.0.5 was interpreted literally, in the case of the Inservice Testing Program, it would require the licensee to address these situations by shutting the plant down to perform testing.

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, U.S. Nuclear Regulatory Commission B15482/ Attachment 3/Page 3 February 5, 1996 As stated in NUREG-1482, the NRC recommends that licensees revise the technical specifications to include the recommendations from the revised standard technical specifications (NUREG-1431) for the inservice inspection and testing programs. With the revisions to the technical specifications, upon finding an ASME Code requirement impractical because of prohibitive dose rates or limitations in the design, construction, or system configuration, the licensee can implement the relief request at that time, provided the relief request has been 1) acceptably reviewed pursuant to 10CFR50.59, and

2) approved by the Plant Operations Review Committee. ,

If an impracticality is determined within the initial interval or within the first 12 months of a new interval, the licensee follows the requirements in 10CFR50.55a(f) (5) (iii) and (iv) or (g)(5)(iii) and (iv). If an impractical requirement is identified during subsequent intervals and not within the first 12 months, the licensee must meet the requirements of 10CFR50.55a(f) (5) (iii) or (g) (5) (iii) , notify the Commission, submit the information ,

supporting the determination of impracticality, and obtain NRC approval pursuant to (f) (6) (i) or (g) (6) (i), prior to the time that the next test or inspection is required.

NUREG-1431, " Standard Technical Specifications -

Westinghouse Plants," reflects the NRC Staff's position that a licensee may establish and implement the Inservice Inspection and Inservice Testing Programs in accordance with 10CFR50.55a, and does not require that relief requests be granted before they are implemented. Rather, according to the NRC Staff, 10CFR50.55a(f) (5) (iv) and 10CFR50.55a(g) (5) (iv) allow a licensee up to a full year after the start of a new interval to inform the NRC ,

of those new code requirements which cannot be net and to request '

I a relief.

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. Docket No. 50-423 B15482 L

Attachment 4 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Inservice Inspection and Testing Program Safety Assessment and Significant Hazards Consideration l

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i February 1996

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. U.S. Nuclear Regulatory Commission B15482/ Attachment 4/Page 1 February 5, 1996 Nillstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Inservice Inspection and Testing Program Safety Assessment and Significant Hasards Consideration safety Assessment The Commission's final policy statement on technical specification improvement defines the scope of the technical specifications and provides guidance for technical design items to be included in, or relocated out of, the technical specifications document. On July 19, 1995 (60FR36953), the NRC published the final rule governing the implementation of this policy via a revision of 10CFR50.36, 4 " Technical Specifications" which became offactive August 18, 1995.

NUREG-1431, Revision 1, " Standard Technical Specifications -

Westinghouse Plants," relocates the IST requirements to the administrative section of the technical specifications and deletes a portion of the ISI requirements, retaining the reactor coolant pump (RCP) flywheel inspections in the administrative control section. NUREG-1482, Chapter 6 addresses the situation in which the technical specifications are in conflict with the regulation of 10CFR50.55a and recommends that licensees revise their technical specifications to incorporate the revised standard technical specifications for IST' programs. The second 10-year interval for the Millstone Unit No. 3 IST and ISI programs will begin April 23, 1996, and October 23, 1996, and these IST and ISI programs will be based on the requirements of the 1989 Edition of the ASME Code.

With the revised technical specifications, upon finding a code requirement impractical because of limitations in the design, construction, or system configurations, NNECO would be required to prepare the determination describing the impractical conditions and the applicable code requirements that cannot be met in accordance  !

with 10CFR50.55a, paragraphs f(5) (iii) and f(5) (iv), and g(5) (iii) and g (5) (iv) if within the first 12 months of a new interval.

Specifically, 10CFR50.55a(f)(iv) and 10CFR50. 55a (g) (iv) allow a licensee up to a full year after a beginning of the updated interval to inform the NRC of those new Code requirements which cannot be met and to request relief. Therefore, with the revision j to the technical specifications, upon finding an ASME Code  ;

requirement impractical because of prohibitive dose rates or limitations in the design, construction or system configuration, NNECO can implement the relief request at that time. This implementation could occur provided the relief request has been l (1) acceptably reviewed pursuant to 10CFR50.59, and (2) approved by l the Plant Operations Review Committee. I Although NUREG-1482 does not specifically address the ISI program, the situation is applicable to both the ISI and IST programs. By

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[ U.S. Nuclear Regulatory Commission B15482/ Attachment 4/Page 2 February 5, 1996 I

rulemaking effective September 8, 1993 (57FR34666), the NRC established paragraph (f) to separate the IST program requirements from the ISI requirements in paragraph (g) of 10CFR50.55a. By deleting "g," the reference to 10CFR50.55a implies both "(f)" and

"(g)" requirements and are applicable as appropriate. Therefore, reference to paragraph "(g)" of 10CFR50.55a from Technical Specification 4.0.5a and Bases section 3/4.4.10 is appropriate and acceptable.

In summary, the proposed changes are consistent with the intent of '

the revised standard technical specifications (i.e., NUREG-1431) and the regulatory guidance in NUREG-1482. The ISI and IST requirements are given in 10CFR50.55a, which the licensee documents via its 10-year interval program requirements. The proposed changes eliminate inconsistencies between the technical specifications and the regulations. The proposed changes do not impact the consequences of an accident previously evaluated or reduce the margin of safety as defined in the basis for any technical specifications. Therefore, the proposed changes do not adversely affect or endanger the health or safety of the general public.

The changes to Bases Sections 3/4.7.7 and 3/4.7.8 add design basis information and provide clarification of system design and operation to facilitate interpretation of the technical specification action statements. These action statements cover plant operation when one or both trains of control room envelope pressurization are inoperable. These proposed changes only affect the Bases Section, therefore, they were reviewed in accordance with the provisions of 10CFR50.59 and approved by the Millstone Unit No. 3 Plant Operations Review Committee. These changes are being )

submitted to the NRC for information only. j Significant Hazards consideration Pursuant to 10CFR50.92, NNECO has reviewed the proposed changes to Technical Specification 4.0.5a and Bases Section 3/4.4.10 and has concluded that the changes do not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92 (c) are not compromised. The proposed ,

changes do not involve an SHC because the changes would not: 1

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. ,

1 The proposed changes would remove the wording "...(g), except I where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g) (6) (1) ." The Inservice Inspection and Testing l

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4

U.S. Nuclear Regulatory Commission j B15482/ Attachment 4/Page 3 February 5, 1996 .
Programs are described in the technical specifications pursuant .

l to 10CFR50.55a. In addition, the proposed changes, in .

. accordance with NUREG-1431 and NUREG-1482, would provide relief  !

! to the ASME Code requirement in the interim between the time  !

l of submittal of a relief request until the NRC has issued a 1 safety evaluation and granted the relief. The changes being proposed are administrative in nature and do not affect  :

assumptions contained in plant safety analyses, the physical

! design and/or operation of the plant, nor do they affect any  ;

technical specification that preserves safety analysis

assumptions. Any relief from the approved ASME Section XI Code l requirements will require a 10CFR50.59 evaluation to ensure no technical specification changes or unreviewed safety questions i exist. Therefore, operation of the facility in accordance with l the proposed changes would not affect the probability or I l

consequences of an accident previously analyzed.

2. Create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes would remove the wording ". . . (g) , except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section

50. 55a (g) (6) (i) . " The Inservice Inspection and Testing Programs are described in the technical specifications pursuant to 10CFR50.55a. In addition, the proposed changes, in accordance with NUREG-1431 and NUREG-1482, would provide relief to the ASME Code requirement in the interim between the time )

of submittal of a relief request until the NRC has issued a i safety evaluation and granted relief. The changes being  !

proposed are administrative in nature and will not change the physical plant or the modes of operation defined in the facility license. The changes do not involve the addition or modification of equipment nor do they alter the design or operation of plant systems. Any relief from the approved ASME Section XI Code requirements will require a 10CFR50.59 evaluation to ensure no technical specification changes or unreviewed safety questions exist. Therefore, operation of the facility in accordance with the proposed changes would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in the margin of safety. ,

The proposed changes would remove the wording "...(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g) (6) (i) . " The Inservice Inspection and Testing Programs are described in the technical specifications pursuant

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'," U.S. Nuclear Regulatory Commission B15482/ Attachment 4/Page 4 February 5, 1996 I to 10CFR50.55a. In addition, the proposed changes, in )

accordance with NUREG-1431 and NUREG-1482, would provide relief '

to the ASME Code requirement in the interim between the time of submittal of a relief request until the NRC has issued a safety evaluation and granted the relief. The changes being proposed are administrative in nature and will not alter the bases for assurance that safety-related activities are performed correctly or the basis for any technical specification that is related to the establishment or l maintenance of a safety margin. Any relief from the approved )

ASME Section XI Code requirements will require a 10CFR50.59 evaluation to ensure no technical specification changes or unreviewed safety questions exist. Therefore, operation of the facility in accordance with the proposed changes would not involve a significant reduction in a margin of safety.

The Commission has provided guidance concerning the application of the standards of 10CFR50.92 by providing certain examples (51FR7751, March 6, 1986) of amendments that are not considered likely to involve an SHC. The changes proposed herein are enveloped by example (vii), a change to conform a license to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations. The proposed changes eliminate inconsistencies )

between the technical specifications and the regulations, are consistent with the guidance in NUREG-1431 and NUREG-1482, and a recently issued amendment on the Wolf Creek Nuclear Operating Corporation docket (docket no. 50-482). As previously stated, the proposed changes do not involve an SHC.

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