ML021290137

Technical Specifications for Amendments Relocating Various Reactor Coolant System Technical Specifications to the Respective Units Technical Requirements Manual

ML021290137
Person / Time
Site:	Millstone
Issue date:	05/08/2002
From:	NRC/NRR/DLPM
To:
Nerses V, NRR/DLPM, 415-1484
Shared Package
ML021330111	List: ML020870765 ML021290137
References
TAC MB2240, TAC MB2273
Download: ML021290137 (29)

Text

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE S..

3/4 4-2 S..

3/4 4-3 SECTION 3/4.4.2 3/4.4.3 3/4.4.4 3/4.4.5 3/4.4.6 3/4.4.7 3/4.4.8 3/4.4.9 3/4.4. 10 3/4.4.11 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 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS 3/4.5.2 ECCS SUBSYSTEMS Tavg Ž 300°F 3/4.5.3 ECCS SUBSYSTEMS Tavg< 300°F 3/4 5.4 REFUELING WATER STORAGE TANK TRISODIUM PHOSPHATE (TSP) 3/4 3/4 3/4 3/4 3/4 5-9 MILLSTONE -

UNIT 2 0801 VI Amendment No.

0, 77, 10f, 197, 777, 79, 266 SAFETY VALVES.............

RELIEF VALVES.............

PRESSURIZER...............

STEAM GENERATORS REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems Reactor Coolant System Leakage DELETED..................

SPECIFIC ACTIVITY..........

PRESSURE/TEMPERATURE LIMITS Reactor Coolant System DELETED Overpressure Protection Systems DELETED DELETED..................

4-4 4-5 4-8 4-8 4-9 4-10 4-13 4-17 4-17 4-21 4-21a 4-22 4-23 5-1 5-3 5-7 5-8 3/4 5.5 I

INDEX BASES SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.2 SAFETY VALVES 3/4.4.3 RELIEF VALVES..................

3/4.4.4 PRESSURIZER 3/4.4.5 STEAM GENERATORS 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY..............

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

3/4.4.10 DELETED 3/4.4.11 DELETED.....

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK (RWST) 3/4.5.5 TRISODIUM PHOSPHATE (TSP).......

PAGE B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a B 3/4 4-3 B 3/4 4-4 B 3/4 4-4 B 3/4 4-5 B 3/4 4-7 B 3/4 4-8 B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-3 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 SECONDARY CONTAINMENT......

MILLSTONE - UNIT 2 0802 B 3/4 6-1 B 3/4 6-3 B 3/4 6-3 B 3/4 6-4 B 3/4 6-5 XII Amendment No. f, 9, 77, 109, M, 709, 577, 799, 266 i

This page intentionally left blank MILLSTONE - UNIT 2 0720 3/4 4-10 Amendment No. 266 I

This page intentionally left blank Amendment No. 266 MILLSTONE - UNIT 2 0720 I

3/4 4-11

This page intentionally left blank Amendment No. 266 1 MILLSTONE - UNIT 2 0720 3/4 4-12

This page intentionally left blank Amendment No.

?IF, 266 1 MILLSTONE - UNIT 2 0721 3/4 4-21

This page intentionally left blank MILLSTONE -

UNIT 2 0722 3/4 4-23 Amendment No. 70, 77, 179, 7Lý,

7ýX, 770, 266 I

REACTOR COOLANT SYSTEM BASES 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uCi/gram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

MILLSTONE - UNIT 2 B 3/4 4-4 Amendment No.,1X, J}, 266 0723

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°F 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 120°F 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 (161IF 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 700 F, whenever any reactor vessel stud is tensioned.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. Removal of reactor vessel irradiation surveillance specimens does not constitute a CORE ALTERATION per Specification 1.12.

MILLSTONE - UNIT 2 B 3/4 4-7 Amendment No. 9, 79, gf, 11p, 266 0724

BASES 3/4.4.11 DELETED Amendment No. 110, 770, 266 MILLSTONE - UNIT 2 0725 B 3/4 4-8

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION TABLE 3.3-13 DELETED TABLE 4.3-9 DELETED 3/4.3.4 DELETED 3/4.3.5 SHUTDOWN MARGIN MONITOR.....

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.....

Hot Standby...........

Hot Shutdown Cold Shutdown

- Loops Filled Cold Shutdown

- Loops Not Filled Isolated Loop............

Isolated Loop Startup......

3/4.4.2 SAFETY VALVES..........

DELETED............

3/4.4.3 PRESSURIZER Startup and Power Operation..................

FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL..................

Hot Standby..........

3/4.4.4 RELIEF VALVES............

3/4.4.5 STEAM GENERATORS TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.....

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.....

Operational Leakage......

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES 3/4.4.7 DELETED............

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS 3/4.4.8 SPECIFIC ACTIVITY.

MILLSTONE -

UNIT 3 0899 vii 3/4 4-1 3/4 4-2 3/4 4-3 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-8

. 3/4 4-9

. 3/4 4-10

. 3/4 4-11

. 3/4 4-11a

. 3/4 4-1ib

. 3/4 4-12

. 3/4 4-14

. 3/4 4-19

. 3/4 4-20

. 3/4 4-21

. 3/4 4-22 3/4 4-24

. 3/4 4-25

. 3/4 4-26 3/4 4-27 3/4 4-28 Amendment No.

lo, lg#, Xg, 19l, 197, 204 PAGE S.........

3/4 3-82 I

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 >1tCi/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 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE 3/4 4-30

. 3/4 4-31

. 3/4 4-33 3/4 4-34

. 3/4 4-35

. 3/4 4-36 Pressurizer........

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

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

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

3/4.5.2

__ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350'F 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F 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 Containment Integrity Containment Leakage Containment Air Locks Containment Pressure........

. 3/4 5-1 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 ML T7, UA, Mn, 2041 MILLSTONE -

UNIT 3 0786 PAGE THE COLD THE COLD 3/4 4-37 3/4 4-38 3/4 4-40 3/4 4-41 3/4 4-42 3/4 4-43 I I viii Amendment No.

INDEX BASES SECTION 3/4.0 APPLICABILITY 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL.......

3/4.1.2 DELETED 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 3/4.2.4 QUADRANT POWER TILT RATIO 3/4.2.5 DNB PARAMETERS..........

3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 3/4.3.3 3/4.3.4 3/4.3.5 INSTRUMENTATION MONITORING INSTRUMENTATION...

TURBINE OVERSPEED PROTECTION..

SHUTDOWN MARGIN MONITOR 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.2 SAFETY VALVES 3/4.4.3 PRESSURIZER 3/4.4.4 RELIEF VALVES 3/4.4.5 STEAM GENERATORS.....

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE........

3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY 3/4.4.9 PRESSURE/TEMPERATURE LIMITS MILLSTONE - UNIT 3 0900 xiii B 3/4 4-1 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2b B 3/4 4-3 B 3/4 4-4 B 3/4 4-5 B 3/4 4-5 B 3/4 4-7 Amendment No.

fo, $p, IJg, If;,

797, 204 PAGE B 3/4 0-1 B 3/4 1-I B 3/4 1-2 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-3 B 3/4 2-5 B 3/4 2-5 B 3/4 3-1 B 3/4 3-3 B 3/4 3-6 B 3/4 3-7 I

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

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

3/4.4.10 DELETED........

3/4.4.11 DELETED...........

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 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION 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 FLOOD PROTECTION 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 PAGE B 3/4 4-9

. B 3/4 4-10 B 3/4 4-15 B 3/4 4-15

. B 3/4 5-1 B 3/4 5-1 B 3/4 5-2

. B 3/4 5-3

. B 3/4 6-1

. B 3/4 6-2

. B 3/4 6-3

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

. B 3/4 6-4 B 3/4 7-1 B 3/4 7-7 B 3/4 7-7 B 3/4 7-7 B 3/4 7-8 3/4 7-10 3/4 7-10 3/4 7-17 3/4 7-23 3/4 7-23 xiv Amendment No.

fg, FP,

119, 119, 779, 204 MILLSTONE -

UNIT 3 0786

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES AND PROGRAMS.............

6-14 6.9 REPORTING REQUIREMENTS............

.. 6-17 6.9.1 ROUTINE REPORTS................

6-17 Startup Report...............

6-17 Annual Reports...............

6-18 Annual Radiological Environmental Operating Report..........

6-19 Annual Radioactive Effluent Release Report..................

6-19 Monthly Operating Reports............

6-19 Core Operating Limits Report.........................

6-19a 6.9.2 SPECIAL REPORTS................

6-21 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM 6.12 HIGH RADIATION AREA.......

6.13 RADIOtOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) 6.14 RADIOACTIVE WASTE TREATMENT..................

6.15 RADIOACTIVE EFFLUENT CONTROLS PROGRAM..........

6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM...

6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM MILLSTONE - UNIT 3 xix 0787 6-22 6-23 S....

. 6-24 S....

. 6-24 S....

. 6-25 S....

. 6-26 S....

. 6-26 Amendment No.

A, M 99, J77,

9, 204

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204

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0ILLSTONE - UNIT 3 0788 3/4 4-27

This page intentionally left blank Amendment No.

197, 204 1

MILLSTONE - UNIT 3 0789 3/4 4-37

This page intentionally left blank Amendment No.

97, ;p, 204

,I MILLSTONE - UNIT 3 0790 3/4 4-42

This page intentionally left blank Amendment No. 9, 79, 204 MILLSTONE -

UNIT 3 0790 3/4 4-43

This page intentionally left blank Amendment No. 79, F, 17, 204 MILLSTONE - UNIT 3 0790 3/4 4-43a

ADMINISTRATIVE CONTROLS

j.

Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

6.16 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

The program shall provided (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the REMODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

a.

Monitoring,

sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the REMODCM.

b.

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

c.

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.17 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM This program shall provide for the inspection of each reactor coolant pump flywheel by either qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (magnetic partical testing and/or penetrant testing) of exposed surfaces defined by the volume of the disas-sembled flywheels at least once every 10 years.

Amendment No. J9, 204 MILLSTONE -

UNIT 3 0791 6-26

REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

The specified allowable leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure.

It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.

Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady state reactor-to-secondary steam generator leakage rate of 1 gpm.

The values MILLSTONE -

UNIT 3 0792 Amendment No. 204 1

B 3/4 4-5

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.

The Surveillance Requirement to remove and examine the reactor vessel material irradiation surveillance specimens is in accordance with the requirements of 10CFR50, Appendix H.

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.

ASTM 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.

Amendment No.

H, 197, Jf7, 204 1

MILLSTONE -

UNIT 3 0901 B 3/4 4-12

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1. P, 7P7, 204 0ILLSTONE -

UNIT 3 0793 B 3/4 4-13

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OF, 9, 107, 204 1

MILLSTONE - UNIT 3 0793 B 3/4 4-14

REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS BACKGROUND The Cold Overpressure Protection System limits RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the isothermal beltline pressure and temperature (P/T) limits developed using the guidance of ASME Section XI, Appendix G (Reference 1) as modified by ASME Code Case N-640 (Reference 2).

The reactor vessel is the limiting RCPB component for demonstrating such protection.

Cold Overpressure Protection consists of two PORVs with nominal lift setting as specified in Figures 3.4-4a and 3.4-4b, or two residual heat removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and an RCS vent of sufficient size.

Two relief valves are required for redundancy.

One relief valve has adequate relieving capability to prevent overpressurization of the RCS for the required mass input capability.

MILLSTONE - UNIT 3 0902 B 3/4 4-15 Amendment No. 99, 99, 7J9, 197,

197, 204 I

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77, 204 I