ML030060083

Units, 2 and 3, Technical Specification Pages for Amendments 272 & 214 Relocate Selected Millstone Units 2 and 3 Technical Specification Related to the Reactor Coolant System Plant Systems to the Technical Requirements Manual

ML030060083
Person / Time
Site:	Millstone
Issue date:	01/02/2003
From:	NRC/NRR/DLPM/LPD1
To:
References
TAC MB4066, TAC MB4067
Download: ML030060083 (28)

Text

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE......

Safety Valves......

Auxiliary Feedwater Pumps..............

Condensate Storage Tank................

Activity Main Steam Line Isolation Valves Main Feedwater Isolation Components (MFICs)

Atmospheric Dump Valves................

Steam Generator Blowdown Isolation Valves..

3/4.7.2 DELETED 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 3/4.7.4 SERVICE WATER SYSTEM 3/4.7.5 DELETED........

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM..

3/4.7.7 DELETED............................

3/4.7.8 SNUBBERS 3/4.7.9 DELETED............................

3/4.7.10 DELETED............................

3/4.7.11 ULTIMATE HEAT SINK 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating.....

Shutdown 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS.

A.C. Distribution - Operating.

A.C. Distribution - Shutdown D.C. Distribution - Operating.

D.C. Distribution - Shutdown D.C. Distribution (Turbine Battery)

MILLSTONE - UNIT 2 VIII 0894 f

27, PAGE 3/4 7-1 S..

3/4 7-1 S....

3/4 7-4 S.....

. 3/4 7-6 S.....

. 3/4 7-7 S...... 3/4 7-9 3/4 7-9a

... 3/4 7-9c 3/4 7-9d 3/47-10 3/4 7-11 3/4 7-12 3/47-13 3/4 7-16 3/4 7-19 3/4 7-21 3/4 7-33 3/4 7-33 3/4 7-34 3/4 8-1 3/4 8-1 3/4 8-5 3/4 8-6 3/4 8-6 3/4 8-7 3/4 8-8 3/4 8-10 3/4 8-11 Operating Amendment No. 77, 0, I, 190,

, ll, F 1,17, Ml, M. M.

INDEX BASES SECTION 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE......

3/4.7.2 DELETED 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 3/4.7.4 SERVICE WATER SYSTEM 3/4.7.5 DELETED........

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 3/4.7.7 DELETED 3/4.7.8 SNUBBERS 3/4.7.9 DELETED........

3/4.7.10 DELETED........

3/4.7.11 ULTIMATE HEAT SINK 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.................

3/4.9.2 INSTRUMENTATION......

3/4.9.3 DECAY TIME 3/4.9.4 CONTAINMENT PENETRATIONS 3/4.9.5 DELETED...........................

3/4.9.6 DELETED........

3/4.9.7 DELETED........

3/4.9.8 SHUTDOWN COOLING AND COOLING RECIRCULATION MILLSTONE - UNIT 2 0895 XIII Amendment No. ?, 77, M Y, Jf, Iff, Xfl, If?, 272 PAGE B 3/4 7-1 B 3/4 7-3 B 3/4 7-3 B 3/4 7-4 B 3/4 7-4 B 3/4 7-4 B 3/4 7-5 B 3/4 7-5 B 3/4 7-6 B 3/4 7-7 B 3/4 7-7 B 3/4 8-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-2 B 3/4 9-2 B 3/4 9-2 I

I

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1

a.

The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b.

DELETED Amendment No. ?JIF, 272 M7LLSTONE - UNIT 2 0774 3/4 4-18

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 0775 Amendment No. If, 272 3/4 4-20

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 0776 3/4 7-10 Amendment No. 2 7 2

THIS PAGE INTENTIONALLY LEFT BLANK 1Amendment No. 19, 2 7 2 77LLSTONE - UNIT 2 0777 3/4 7-13

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 0777 3/4 7-14 Amendment No.

• 77, 272

THIS PAGE-INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 2 0777 3/4 7-15 Amendment Ho. 272

THIS PAGE INTENTIONALLY LEFT BLANK Amendment No. -J J,

91*, 2 7 2 7ILLSTONE - UNIT 2 0778 3/4 7-19

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE -

UNIT 2 0778 3/4 7-20 Amendment Noo. 2 7 2

REACTOR COOLANT SYSTEM BASES The maximum RTNDT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 50°F.

The Lowest Service Temperature limit is based upon this RTNDT since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100 0F for piping, pumps and valves.

Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.

Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement.

Included in this evaluation is consideration of flange protection in accordance with 10 CFR 50, Appendix G.

The requirement makes the minimum temperature RTNDT plus 90°F for hydrostatic test and RTNDT plus 1200F for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure.

Since the flange region RTNDT has been calculated to be 30°F, the minimum flange pressurization temperature during normal operation is 150°F (161°F with instrument uncertainty) when the pressure exceeds 20% of the preservice hydrostatic pressure.

Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement.

To establish the minimum boltup temperature, ASME Code Section XI, Appendix G, requires the temperature of the flange and adjacent shell and head regions shall be above the limiting RTNDT temperature for the most limiting material of these regions.

The RTNDT temperature for that material is 30°F.

Adding 10.5°F, for temperature measurement uncertainty, results in a minimum boltup temperature of 40.5°F.

For additional conservatism, a minimum boltup temperature of 70°F is specified on the heatup and cooldown curves.

The head and vessel flange region temperature must be greater than 70°F, whenever any reactor vessel stud is tensioned.

MILLSTONE - UNIT 2 B 3/4 4-7 Amendment No. Y?, 70, Y, 719, 0897

,272

rLM11II

.JI Ja!L-'i BASES a feedwater isolation signal since the steam line break accident analysis credits them in prevention of feed line volume flashing in some cases.

Feedwater pumps are assumed to trip immediately with an MSI signal.

3/4.7.1.7 ATMOSPHERIC DUMP VALVES The atmospheric dump valve (ADV) lines provide a method to maintain the unit in HOT STANDBY, and to replace or supplement the condenser steam dump valves to cool the unit to Shutdown Cooling (SDC) entry conditions.

Each ADV line contains an air operated ADV, and an upstream manual isolation valve.

The manual isolation valves are normally open, and the ADVs closed.

The ADVs, which are normally operated from the main control

room, can be operated locally using a manual handwheel.

An ADV line is OPERABLE if local manual operation of the associated valves can be used to perform a controlled release of steam to the atmosphere.

This is consistent with the LOCA analysis which credits local manual operation of the ADV lines for accident mitigation.

3/4.7.1.8 STEAM GENERATOR BLOWDOWN ISOLATION VALVES The steam generator blowdown isolation valves will isolate steam generator blowdown on low steam generator water level.

An auxiliary feedwater actuation signal will also be generated at this steam generator water level.

Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main feedwater.

The steam generator blowdown isolation valves will also close automatically upon receipt of a containment isolation signal or a high radiation signal (steam generator blowdown or condenser air ejector discharge).

3/4.7.2 DELETED 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM The OPERABILITY of the Reactor Building Closed Cooling Water (RBCCW)

System ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions.

The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

The RBCCW loops are redundant of each other to the degree that each has separate controls and power supplies and the operation of one does not depend on the other.

In the event of a design basis accident, one RBCCW loop is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water.

To ensure this requirement is

met, two RBCCW loops must be

OPERABLE, and independent to the extent necessary to ensure that a single failure will not result in the unavailability MILLSTONE - UNIT 2 B 3/4 7-3a Amendment No.

7, fl,

11f, 0796 710, 272

PLANT SYSTEMS BASES 3/4.7.4 SERVICE WATER SYSTEM (Continued)

The Technical Specification Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the accident analysis are met and that subsystem OPERABILITY is maintained.

The purpose of the service water pumps differential pressure test, Surveillance Requirement 4.7.4.1.a.2, a substantial flow test, is to ensure that the pumps have not degraded to a point where the accident analysis would be adversely impacted.

The surveillance requirement acceptance criteria for the service water pumps was developed assuming a 7% degraded pump from the actual pump curves.

Flow and pressure measurement instrument inaccuracies for the service water pumps have been accounted for in the design basis hydraulic analysis.

It is not necessary to account for flow and pressure measurement instrument inaccuracies in the acceptance criteria contained in the surveillance procedure.

3/4.7.5 DELETED 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 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 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",

10 CFR 50.

The LCO is modified by a footnote allowing the control room boundary to be opened intermittently under administrative controls.

For entry and exit through doors the administrative control of the opening is performed by the person(s) entering or exiting the area.

For other openings, these controls consist of stationing a dedicated individual at the opening who is in constant communication with the control room.

This individual will have a method to rapidly close the opening when a need for control room isolation is indicated.

The control room radiological dose calculations use the conservative minimum acceptable flow of 2250 cfm based on the flowrate surveillance requirement of 2500 cfm + 10%.

Amendment No.

17, ?I, 272 8ILLSTONE - UNIT 2 0898 B 3/4 7-4a

PLANT SYSTEMS BASES 3/4.7.7 DELETED 3/4.7.8 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

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 expected 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 Guide 8.8 and 8.10.

The addition or deletion of any hydraulic or mechanic 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 systems.

Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection.

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.

MILLSTONE - UNIT 2 B 3/4 7-5 Amendment Nos. X7, 1, Y, XXf, 0896 70, 272

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >lgCi/gram DOSE EQUIVALENT 1-131........

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...........

3/4.4.9 PRESSURE/TEMPERATURE LIMITS.................

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 10 EFPY FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 10 EFPY TABLE 4.4-5 DELETED Pressurizer..........

Overpressure Protection Systems...............

FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE OVERPRESSURE SYSTEM (FOUR LOOP OPERATION)

FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE OVERPRESSURE SYSTEM (THREE LOOP OPERATION) 3/4.4.10 DETETED 3/4.4.11 DELETED 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............

3/4.5.2 3/4.5.3 3/4.5.4

. 3/4 4-33

. 3/4 4-34

. 3/4 4-35

. 3/4 4-36

. 3/4 4-37 S...

. 3/4 4-38 COLD COLD

. 3/4 4-40 3/4 4-41

• 3/4 4-42

• 3/4 4-43

......... 3/4 5-1 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F ECCS SUBSYSTEMS - Tavg LESS THAN 350°F REFUELING WATER STORAGE TANK.....

3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS.........

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity Containment Leakage Containment Air Locks Containment Pressure..................

3/4 5-3 3/4 5-7

• 3/4 5-9 3/4 5-10 3/4 6-1 3/4 6-2

• 3/4 6-5 3/4 6-7 viii MILLSTONE - UNIT 3 0870 Amendment No. Ay, 07.F, 214 700214 PAGE 3/4 4-30

. 3/4 4-31 I

Q 0

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 TABLE 4.7-1 3/4.7.2 3/4.7.3 3/4.7.4 3/4.7.5 3/4.7.6 3/4.7.7 3/4.7.8 3/4.7.9 3/4.7.10 TABLE 4.7-2 FIGURE 4.7-1 3/4.7.11 3/4.7.12 Table 3.7-4 Table 3.7-5 3/4.7.13 3/4.7.14 TABLE 3.7-6 STEAM LINE SAFETY VALVES PER LOOP Auxiliary Feedwater System Demineralized Water Storage Tank Specific Activity SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM Main Steam Line Isolation Valves Steam Generator Atmospheric Relief Bypass Lines DELETED REACTOR PLANT COMPONENT COOLING WATER SYSTEM SERVICE WATER SYSTEM ULTIMATE HEAT SINK DELETED CONTROL ROOM EMERGENCY VENTILATION SYSTEM CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM AUXILIARY BUILDING FILTER SYSTEM SNUBBERS SNUBBER VISUAL INSPECTION INTERVAL SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST DELETED DELETED DELETED DELETED DELETED AREA TEMPERATURE MONITORING AREA TEMPERATURE MONITORING 3/4 7-3 3/4 7-4 3/4 7-6 3/'4 7-7 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 7-8 7-9 7-9a 7-10 7-11 7-12 7-13 7-14 7-15 7-18 7-20 7-22 7-27 7-29 7-30 x

MILLSTONE - UNIT 3 0952 Amendment No. fl, f, IP, If, 214 3/4 7-32 3/4 7-33 I

INDEX BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>IMeV)

AS A FUNCTION OF FULL POWER SERVICE LIFE.....................

... B 3/4 4-10 3/4.4.10 DELETED...........

... B 3/4 4-15 3/4.4.11 DELETED.........

.... B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.........

3/4.5.4 REFUELING WATER STORAGE TANK 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT...............

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4.6.4 COMBUSTIBLE GAS CONTROL......

3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.6 SECONDARY CONTAINMENT.............

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE......

3/4.7.2 DELETED........

3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM 3/4.7.4 SERVICE WATER SYSTEM 3/4.7.5 ULTIMATE HEAT SINK 3/4.7.6 DELETED........

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.

3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM.

3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM 3/4.7.10 SNUBBERS MILLSTONE - UNIT 3 xiv 0953 S....

. B 3/4 5-1 S........

B 3/4 5-1 S....

. B 3/4 5-2 S....

. B 3/4 5-3 S....

. B 3/4 6-1 S....

. B 3/4 6-2 S.......

. B 3/4 6-3 S....

. B 3/4 6-3 S....

. B 3/4 6-3b S....

. B 3/4 6-4 S......

B 3/4 7-1 S....

. B 3/4 7-7 S....

. B 3/4 7-7 S....

. B 3/4 7-7 S....

. B 3/4 7-8 S......

B 3/4 7-10 S......

B 3/4 7-10 S......

B 3/4 7-17 S......

B 3/4 7-23 S......

B 3/4 7-23 Amendment No. f? g, 17f, 214 BASES I

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INDEX BASES SECTION 3/4.7.11 GELETED 3/4.7.12 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C.

SOURCES, D.C.

ONSITE POWER DISTRIBUTION 3/4.8.4 DELETED.....

PAGE

. B 3/4 7-25 S....

. B 3/4 7-25

SOURCES, AND B 3/4 8-1 B 3/4 8-3 3/4.9-REFUELING OPERATIONS 3/4.9.1 3/4.9.2 3/4.9.3 3/4.9.4 3/4.9.5 3/4.9.6 3/4.9.7 3/4.9.8 3/4.9.9 3/4.9. 10 3/4.9.12 3/4.9.13 3/4.9.14 3/4.*0 o 3/4.10.1 3/4.10.2 3/4.10.3 3/4.!0.4 BORON CONCENTRATION INSTRUMENTATION DECAY TIME CONTAINMENT BUILDING PENETRATIONS COMMUNICATIONS REFUELING MACHINE CRANE TRAVEL SPENT FUEL STORAGE AREAS.........

RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL FUEL BUILDING EXHAUST FILTER SYSTEM...........

SPENT FUEL POOL -

REACTIVITY SPENT FUEL POOL -

STORAGE PATTERN SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN GROUP HEIGHT, INSERTiON, AND POWER DISTRIBUTION LIMITS PHYSICS TESTS......................

REACTOR COOLANT LOOPS..................

3/4.10.5 DELETED MILLSTONE -

UNIT 3 0953 xv Amendment No.

F;, p, 100, 797, ;19, 17P. IFY 19 797, 214 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 9-1 9-I 9-I 9-1 9-1 9-2 9-2 9-2 9-7 9-8 9-8 9-8 9-9 10-1 10-1 10-1 10-1 10-1 i

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XLACIOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.]

Reactor Coolant System (except the pressurizer) temperature, pressure, and heatup and cooldown rates of ferritic materials shall be limited in accordance with the limits shown on Figures 3.4-2 and 3.4-3.

In addition, a maximum of one reactor coolant pump can be in operation when the lowest unisolated Reactor Coolant System loop wide range cold leg temperature is

< 160"F.

APPLICABILITY:

At all times.

ACTION:

a.

With any of the above limits exceeded in MODES 1, 2, 3, or 4, perform the following:

1. Restore the temperature and/or pressure to within limit within 30 minutes.

AND

2.

Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System remains acceptable for continued operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Otherwise, be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 with RCS pressure less than 500 psia within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With any of the above limits exceeded in other than MODES 1, 2, 3, or 4, perform the following:

1. Immediately initiate action to restore the temperature and/or pressure to within limit.

AND

2.

Perform an engineering evaluation to determine the effects of the out of limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System is acceptable for continued operation prior to entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup and cooldown operations, and during the one-hour period prior to and during inservice leak and hydrostatic testing operations.

4.4.9.1.2 DELETED Amendment No. 7M7, 777, 214 MILLSTO3 E - UNIT 3 0873 3/4 4-33

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 0874 3/4 4-36 Amendment No. f*,

W7, JP7, 214

THIS PAGE INTENTIONALLY LEFT BLANK Amendment No. 214 MILLSTONE - UNIT 3 0875 3/4 7-10

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 0876 3/4 7-14 Amendment No. jf, 214

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE -

UNIT 3 0877 3/4 7-30 Amendment No.

17, mP, 214

THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE -

UNIT 3 0877 3/4 7-31 Amendment No. 709, 214

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (continued)

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This Surveillance Requirement is only required to be performed during system heatup, cooldown, and ISLH testing.

No Surveillance Requirement is given for criticality operations because LCO 3.1.1.4 contains a

more restrictive requirement.

It is not necessary to perform Surveillance Requirement 4.4.9.1.1 to verify compliance with Figures 3.4-2 and 3.4-3 when the reactor vessel is fully detensioned.

During refueling, with the head fully detensioned or off the reactor vessel, the RCS is not capable of being pressurized to any significant value.

The limiting thermal stresses which could be encountered during this time would be limited to flood-up using RWST water as low as 40"F.

It is not possible to cause crack growth of postulated flaws in the reactor vessel at normal refueling temperatures even injecting 40°F Water.

REFERENCES

1.

ASME Boiler and Pressure Vessel

Code,Section XI, Appendix G, "Fracture Toughness for Protection Against Failure," 1995 Edition.

2.

ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," dated February 26, 1999.

3.

10 CFR 50 Appendix G, "Fracture Toughness Requirements."

4.

AST;; E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706."

5.

10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

6.

Regulatory Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," dated May 1988.

7.

ASME Boiler and Pressure Vessel

Code,Section XI, Appendix E, "Evaluation of Unanticipated Operating Events," 1995 Edition.

MILLSTONE - UNIT 3 B 3/4 4-12 Amendment No. f, 77, fl7, 0878 X,

214

PLANT SYSTEMS BASES 3/4.7.1.6 STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES The OPERABILITY of the steam generator atmospheric relief bypass valve (SGARBV) lines provides a method to recover from a steam generator tube rupture (SGTR) event during which the operator is required to perform a limited cooldown to establish adequate subcooling as a necessary step to limit the primary to secondary break flow into the ruptured steam generator.

The time required to limit the primary to secondary break flow for an SGTR event is more critical than the time required to cooldown to RHR entry conditions.

Because of these time constraints, these valves and associated flow paths must be OPERABLE from the control room.

The number of SGARBVs required to be OPERABLE from the control room to satisfy the SGTR accident analysis requires consideration of single failure criteria.

Four SGARBV are required to be OPERABLE to ensure the credited steam release pathways available to conduct a unit cooldown following a SGTR.

For other design events, the SGARBVs provide a safety grade method for cooling the unit to residual heat removal (RHR) entry conditions should the preferred heat sink via the steam bypass system or the steam generator atmospheric relief valves be unavailable.

Prior to operator action to cooldown, the main steam safety valves (MSSVs) are assumed to operate automatically to relieve steam and maintain the steam generator pressure below design limits.

Each SGARBV line consists of one SGARBV and an associated block valve (main steam atmospheric relief isolation valve, 3MSS*MOV18A/B/C/D).

These block valves are used in the event a steam generator atmospheric relief valve (SGARV) or SGARBV fails to close.

Because of the electrical power relationship between the SGARBV and the block valves, if a block valve is maintained closed, the SGARBV flow path is inoperable because of single failure consideration.

The bases for the required actions can be found in NUREG 1431, Rev.

1.

The LCO APPLICABILITY and ACTION statements uses the terms "MODE 4 when steam generator is relied upon for heat removal" and "in MODE 4 without reliance upon steam generator for heat removal."

This means that those steam generators which are credited for decay heat removal to comply with LCO 3.4.1.3 (Reactor Coolant System, Hot Shutdown) shall have an OPERABLE SGARBV line.

See Bases Section 3/4.4.1 for more detail.

3/4.7.2 DELETED MILLSTONE - UNIT 3 B 3/4 7-7 Amendment No. Jt, J9I, JF1, 214 0879

PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UIS temperature is measured at the locations described in the LCO write-up provided in this section.

Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature.

The 24-hour frequency is based on operating experience related to trending of the parameter variations during the applicable modes.

This surveillance requirement verifies that the average water temperature of the UHS is less than or equal to 75°F.

Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70°F during the applicable modes.

The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70"F.

The frequency is based on operating experience related to trending of the parameter variations during the applicable modes.

3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND 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 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.

MILLSTONE - UNIT 3 B 3/4 7-10 Amendment No.

7, J*, 75, 08*0 214

PLANT SYSTEMS BASES 3/4.7.10 SNUBBERS (Continued)

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 indicatin 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.

3/4.7.11 DELETED 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.

MILLSTONE - UNIT 3 B 3/4 7-25 Amendment Nos. g, F, log, 1y, 0954 214 11,