ML20081K993

Amends 30 to Licenses NPF-72 & NPF-77,respectively.Amend Revises Tech Specs Re Heatup & Cooldown Curves & Power Operated Relief Valve Low Temp Overpressure Protection Setpoints

ML20081K993
Person / Time
Site:	Braidwood
Issue date:	06/27/1991
From:	Barrett R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20081K997	List: ML20081K993 ML20081K999
References
NUDOCS 9107020271
Download: ML20081K993 (25)

Text

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UNITED STAT ES 4

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,.gv g COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50 456 BRAIDWOOD STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 30 License No. NPF-72 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The apnlication for amendment by Conmonwealth Edison Company (the licensee) dated December 19, 1990, complier, with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; P

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's egulations and all applicable requirements have been satisfied.

2.

Accordingly, the 1.ense is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

9107020271 910627 PDR ADOCK 05000456 P

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1 (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendmerit No. 30 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.

The licensee shall operate the facility in accordance with the Technicel Specifications and 1

the Environmental Protection Plan, 3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f

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Rich r-

. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 27,1991 l

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's NUCLEAR REGULATORY COMMISElON n AsmNotoN. o c. rom u....+s COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT NO. 2 At'ENDMENT TO FACILITY OPERATlf;G LICENSE Amendment No. 30 License ho. NPF-77 1.

The Nuclear Regulatory Commission (the Commission) h s found that:

i A.

The application for amendnent by Commonwealth Edison Company (the licensce) dated De:emner 19, 1990, complies with the standards and requirements of the Atomic Et;ergy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (O that the activities authorized oy this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendmer,t is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

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. (2) Technical Specification $

The Technical Specifications contained in Appendix A as revised through Amendment No. 30 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2,1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date if its issuance.

FOR THE NUCLEAR REGULA ORY C0tiMISS10N r /1

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,l' Rich dA. Barrett, Director Project Directorate 111-2 Division of Reactor Projects - Ill/lV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: June 27, 1991

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ATTACHMENT TO LICENSE AMEN 0 MENT NOS. 30 AND 30 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 4

DOCKET NOS. STN 50 456 AND STN 50-457 Y

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. Overleaf pages identified by an asterisk are provided for convenience.

Remove Paaes Insert Paaes Vil Vll Vill Vill

• 3/4 4-31

• 3/4 4-31 3/4 4-32 3/4 4-32 3/4 4-33 3/4 4-33

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3/4 4-34 3/4 4-34 3/4 4-35 3/4 4-35 3/4 4-36 3/4 4-36 3/4 4-39 3/4 4-39 3/4 4-40 3/4 4-40a 3/4 4-40b 3/4 4 41 3/4.4-41 l

B 3/4 4-7 B 3/4 4-7 i

B 3/4 4-8 B 3/4 4-8 B 3/4 4-9 B 3/4 4-9

• B 3/4 4-10

• B 3/4 4-10 B 3/4 4-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-12 l

B 3/4 4-15 B 3/4 4-15 B 3/4 4-16 8 3/4 4-16

• overleaf pages provided for convenience.

I

_ _. ~. _. _. _ _. _. _ _ _ _.., _ _ _ _ _ _. _ _ _ _

3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION....................................

3/4 3-65 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3 69 High Energy Line Break Isolation Sensors.................

3/4 3-73 TABLE 3.3-14 MIGH ENERGY LINE BREAK INSTRUMENTATION...............

3/4 3-74 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................

3/4 3-75 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR C00LtdT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................

3/4 4-1 Hot Standby.............................................

3/4 4-2 Hot Shutdown.............................................

3/4 4-3 Col d Shutdown - Loop s F 111 ed.............................

3/4 4-5 Cold Shutdown - Loops Not Fi11ed.........................

3/4 4-6 Loop Isolation Valves-Operation..........................

3/4 4-7 Loop Isolation Valves-Shutdown............................

3/4 4-8 3/4.4.2 SAFETY VALVES Shutdown...............................................

3/4 4-9 0perating..............................................

3/4 4-10 3/4.4.3 PRESSURIZER.............................................

3/4 4-11 3/4.4.4 RELIEF VALVES............................................

3/4 4-12 3/4.4.5 STEAM GENERATORS.....................

3/4 4-13 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.........................

3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................

3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................

3/4 4-20 Ope ra ti onal Le a kage......................................

3/4 4-21 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4-23 3/4.4.7 CHEMISTRY................................................

3/4 4-24 BRAIDWOOD - UNITS 1 & 2 VII AMENDMENT NO. 30

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...............

3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS......................................

3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY............................

3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

>l pCi/ GRAM DOSE EQUIVALENT I-131..................

3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.......................

3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................

3/4 4-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)......

3/4 4-33 FIGURE 3.4-2b REACT 0's COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)....

3/4 4-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)......

3/4 4-35 FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)......

3/4 4-36 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE......................

3/4 4-37 Pressurizer........................

3/4 4-38 Overpressure Protection Systems..........

3/4 4-39 FIGURE 3.4-4a NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY (Unit 1) 3/4 4-40a FIGURE 3.4-4b NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM (UNIT 2).............

3/4 4-40b 3/4.4.10 STRUCTURAL INTEGRITY......

3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS...

3/4 4-43 BRAIDWOOD - UNITS 1 & 2 VIII AMENDMENT NO. 30

I.

TABLE 4.4-4 (Continued)

TABLE NOTATIONS

1. Until the specific activity of the Reactor Coolant System is restored within its limits.

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

• • A gross radioactivity analysis shall. consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines.

The total specific activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrapolated back to when the sample was taken.

Determination of the contributors to the gross specific activity I

shall be based upon those energy peaks identifiable with a 95% confidence level.

The latest available data may be used for pure beta-emitting radio-nuclides.

• • • A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radio-iodines, which is identified in the reactor coolant.

The specific activities for these individual radionuclides shall be used in the determination of E for the reactor coolant sample.

Determination of the contributors to E shall be based upon these energy peaks identifiable with a 95% confidence level.

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3RAIDWOOD - UNITS 1 & 2 3/4 4-31

I.

1 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100 F in any 1-hour period, b.

A maximum cooldown of 100 F in any 1-hour period, and A maximum temperature change of less than or equal to 10 F in any c.

1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the' temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be 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 reduce the RCS T and pressure to less than 200 F and avg 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIxEMENTS 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, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as requ' red by 10 CFR Part 50, Appendix H, in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2),

and 3.4-4a for Unit 1 (Figure 3.4-4b for Unit 2).

BRAIDWOOD - UNITS 1 & 2 3/4 4-32 AMENDMENT NO. 30 i

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conten UNITS 1 & 2 3/4 4-33 AMENDMENT N0.,30

CURVES APPLICABLE FOR HEATUP RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.

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BRAIDWOOD - UNITS 1 & 2 3/4 4-34 AMENDMENT No. 30

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• applicability has been reduced per Regulatory Guide 1.99 Revision 2 to 12 EFPY.

The calculation to determine applicability utilized actual copper content of 0.05 wt%.

BRAIDWOOD - UNITS 1 & 2 3/4 4-35 AMENDMENT NO. 30

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BRAIDWOOD - UNITS 1 & 2 3/4 4-36 AMENDMENT NO. 30 l

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3. At least'one of the following Overpressure Protection Systems shall i

be OPERABLE:

I a.

Two residual heat removal (RHR) suction relief valves each with a Setpoint of 450 psig i 1%, or b.

Two power-operated relief valves (PORVs) with lift Setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4a for Unit 1 (3.4-4b for Unit 2) or c.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.

APPLICABILITY:

MODES 4 and S, and MODE 6 with the reactor vessel head on.

ACTION:

a.

With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b.

With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, c.

In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

I BRAIDWOOD - UNITS 1 & 2 3/4 4-39 AMENDMENT N0. 30

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FIGURE 3.4-4a NOMINAL PORY PRES $URE RELIEF $ETPOINT YER$US RCS TEMPERATURE FOR THE COLD OVERPRES5URE PROTECTION SYSTEM APPLICABLE UP 7010 EFPY*(UNIT 1)

• applicability has been reduced per Regulatory Guide 1.99 Revision 2 to 4.5 EFPY.

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)

i FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM (UNIT 2)

BRAIDWD0D - UNITS 1 & 2 3/4 4-40b AMENDMENT NO. 30

i REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereaf ter when the PORV is required C?ERABLE; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a.

For RHR suction relief valve RH8708B verify at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves RH8702A and RH8702B are open.

b.

For RHR suction relief valve RH8708A verify at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves RH8701A and RH8701B are open.

c.

Testing pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the vent (s) is being used for overpressure protection.

• Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

BRAIDWOOD - UNITS 1 & 2 3/4 4'41 AMENDMENT NO. 30 l

I I

^

REACTOR C00LANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b) for the service period specified thereon:

a.

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.

Limit lines for cooldown rates betwean those presented may be obtained by interpolation; and b.

Figures 3.4-2a (3.4-2b)-and 3.4-3a (3.4-3b) define limits to assure prevention of non-ductile failure only.

For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2.

These limit lines shall be calculated periodically using methods provided below, 3.

The secondary side of the steam geMrator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F, 4.

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr I

and 200* F/hr respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F, and 5.

System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accor@ nce with the 1973 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel and Code.

BRAIDWOOD - UNITS 1 & 2 8 3/4 4-7 AMENDMENT NO. 30

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of

~

32* effective full power years for Unit 1 (16 effective full power years for Unit 2) of service life.

The 32*EFPY for Unit 1 (16 EFPY for Unit 2) service life period is chosen such that the limiting RT at the 1/4T location in the NDT core region is greater than the RT f the limiting unirradiated material.

NDT The selection of such a limiting RT assures that all components in the NDT Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-la for Unit 1 (Table B 3/4.4-1b for Unit 2).

Reactor operation and resultant fast neutron (E greater than 1 MeV) Irradiation can cause an increase in the RT There-NDT.

fore, an adjusted reference temperature, based upon the fluence, copper content and nickel content of the material in question, can be predicted using l

Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory NDT Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2.

The heatup 1

and cooldown limit curves of Figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b) include predicted adjustments for this shift in RT at the end of 32* EFPY for Unit 1 NDT (16 EFPY for Unit 2) as well as adjustments for possible errors in the pressure and temperature sensing instruments.

Revised heatup and cooldown curves have been generated for Unit 2 in accordance with Regulatory Guide 1.99 Revision 2.

For Unit 1 the curves remain the same.

However, the hpplicability date has been reduced per Regulatory Guide 1.99 Revision 2 to 4.5 EFPY for heatup and 12.0 EFPY for cooldown.

Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available.

Capsules will be removed in accordance with the requirements of

\\STM E185-73 and 10 CFR Part 50, Appendix H.

The surveillance specimen with-drawal schedule is shown in Table 4.4-5.

The lead factor represents the rela-tionship between the fast neutron flux density at the location o.* the capsule and the inner wall of the reactor vessel.

Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.

The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ART NDT NDT for the equivalent capsule radiation exposure.

"For Unit 1 applicability dates have been revised in accordance with Regulatory Guide 1.99 Revision 2, to 4.5 EFPY for heatup and 12.0 ErPY for cooldown.

BRAIDWOOD - UNITS 1 & 2 8 3/4 4-8 AMENDMENT N0. 30

1 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semi elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor opera-tion limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup g

or cooldown cannot be greater than the reference stress intensity factor, K IR' for the metal temperature at that time.

K is obtained from the reference IR fracture toughness curve, defined in Appendix G to the ASME Code.

The KIR curve is given by the equation:

KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)]

(1)

Where:

K is the reference stress intensity factor as a function of the metal IR temperature T and the metal nil-ductility reference temperature RT

Thus, NDT.

the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

CKIM + kit < KIR (2)

Where:

KIM = the stress intensity factor caused by membrane (pressure)

stress, Kyg = the stress intensity factor caused by the thermal gradients, I

BRA:0 WOOD - UNITS 1 & 2 B 3/4 4-9 AMENDMENT NO. 30

i.

e a

REACTOR COOLANT SYSTEM

_ BASES PRESSURE / TEMPERATURE LIMITS (Continued)

K IR = to the RTc nstant provided by the code as a function of temperature relative NDT of the material, C=

2.0 for level A and B service limits, and C=

1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve.

The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, Kgg, for the reference flaw is computed.

From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.

During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.

From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

location is at a higher tem During cooldown, the 1/4T vessel This condition, of course, perature than the fluid adjacent to the vessel ID.

is not true for the steady-state situation.

It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Kgp at the 1/4T location for finite cooldown rates than for steady-state operation.

Furthermore, if conditions exist such that the increase in KIR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-10

h TABLE B 3/4.4-la E;

REACTOR VESSEL TOUGHNESS (UNIT 1)

Average E

Shelf Energy Z

T RT MATERIAL Cu P

NDT NDT

.NMWD*

MWD **

COMPONENT Heat No.

SPEC.

F F*

ft-lbs ft-lbs Closure Head Dome D1398-1 A533B, Cl. 1 66

.009

-30

-30 129 Closure Head Ring 49C1126-1-1 A508, C1. 3

.02

.009

-20

-20 123 N

Closure Head Flange 2030-V-1 A508, Cl. 2

.11

.009

-20

-20 163 Vessel Flange 122N357VA1 A508, C1. 2

.010

-10

-10 106 Inlet Nozzle 21-3257 A508, C1. 2

.09

.008

-20

-20 144 Inlet Nozzle 21-3257 A508, C1. 2

.09

.010

-10

-10 144 Inlet Nozzle 22-3313 A508, C1. 2

.07

.008

-10

-10 130 Inlet Nozzle 22-3313 A508, Cl. 2

.07

.010 0

0 115 Outlet Nozzle 22-3025 A508, Cl. 2

.13

.013

-10

-10 125 w

}

Outlet Nozzle 4-3329 A508, C1. 2

.08

.009

-20

-20 156 w

Outlet Nozzle 4-3383 A508, C1. 2

.08

.008

-20

-20 147 4

Outlet Nozzle 11-5226 A508, Cl. 2

.09

.007

-10

-10 125 Nozzle Shell SP7016 A508 Cl. 2

.04

.008 10 10 155 Upper Shell***

49D383/

A508, Cl. 3

.05

.008(.73)

-30

-30 122 173 l

49C344-1-1 Lower Shell***

490867/

A508, Cl. 3

.03

.007(.73)

-20

-20 135 151 l

49C813-1-1 Bottom Head Ring 49D148-1-1 A508, C1. 3

.05

.008

-50

-50 147 Bottom Head Dome C4882-1 A533B, Cl. 1.14

.010

-20

-20 123 Upper Shell to***

WF-562

.04

.015(.67) 40 40 80 l

Lower Shell Girth Weld

-70

<-10 151 Weld HAZ m

h

• Normal to major working direction.

• • Major working direction.

• • • Calculations per Regulatory Guide 1.99 Revision 2 use the Nickel content shown in parentheses.

l g

?

O

m s'

TABLE B 3/4.4-lb REACTOR VESSEL TOUGHNESS

'8 (UNIT 2)

Average E

Shelf Energy T

RT y

MATERIAL Cu P

NDT NDT NPMD^

PWD^^

• COMPONENT HEAT NO.

SPEC.

F F*

ft-lbs ft-lbs

" Closure Head Dome B9754-1 A5338, C1. 1 T6

.055

-60

-60 151 l

D*

Closure Head Ring 50C478-1-1 A508, C1. 3

.05

.006

-30

-30 128 Closure Head Flange 2031-V-1 A508, C1. 2

.009 20 20 135 N

Vessel Flange 124P455 A508, C1. 2 07

.010 20 20 128 Inlet Nozzle 41-5414 A508, C1. 2

.07

.008

-10

-10 137 Inlet Nozzle' 41-5414 A508, C1. 2

.07

.009

-10

-10 140 Inlet Nozzle 42-5417 A508, C1. 2

.09

.011

-10

-10 122 Inlet Nozzle 42-5417 A508, C1. 2

.09

.009

-10

-10 116 Outlet Nozzle 4-3502 A508, Cl. 2

.09

.012

-10

-10 155 m

Outlet Nozzle 11-5226 A508, C1. 2

.09

.009

-10

-10 116 m} Outlet Nozzle 4-3481 A508, C1. 2

.07

.008

-10

-10 163 Outlet Nozzle 11-5266 A508, C1 2

.09

.010 10 10 117 4 Nozzle Shell SP7056 A508, C1. 2

.04

.005 30 30 115 Upper Shell***

49D963/

A508, C1. 3

.03

.007(.71)

-30

-30 119 147 N

49C904-1-1 Lower Shell***

500102/

A508, C1. 3

.06

.006(.75)

-30

-30 144 168 50C97-1-1 Bottom Head Ring 49D1066-1-1 A508, C1. 3

.07

.008

-30

-30 156 l

Bottom Head Dome D1429-1 A533B, C1. 1.11

.010

-20

-20 120 Upper Shell to***

WF-562

.04

.015(.67) 40 40 80 l

Lower Shell Girth Weld Weld HAZ

-30

-30 145 b

• oM E

• Normal to major working direction.

g

• • Major working direction.

i

• • • Calculations per Regulatory Guide 1.99 Revision 2 use Nickel content shown in parentheses.

l o

i

?-

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

A notch in the cooldown curve of Figure 3.4-3a for Unit 1 (3.4-3b for Unit 2) may be present due to the added constraint on the vessel closure flange given in Appendix G of 10 CFR 50.

This constraint requires that, at pressures f

greater than 20% of the preservice system hydrostatic test pressure, the flange l

regions that are highly stressed by the bolt preload must exceed the RT NDT the material by at least 120'F.

The flange RTNDT + 120'F may impinge on the cooldown curves and therefore the notch is required.

If no notch is present, this indicates that the vessel closure flange region has been determined to l

be not limiting.

1 HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack IR during heatup is lower than the K for the 1/4T crack during steady-state gg conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive i

thermal stresses and different K

's for steady state and finite heatup rates IR do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered.

Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

l The second portion of the heatup analysis concerns the calculation of j

pressure-temperature limitations for the case in which a 1/4T deep outside l

surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present.

These thermal stresses, of course, are dependent on both j

the rate of heatup and the time (or coolant temperature) along the heatup ramp.

Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rcte of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows.

A composite curve is constructed based on a point-by point BRAIDWOOD - UNITS 1 & 2 B 3/4 4-15 AMENDMENT NO. 30

^

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from.the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside

~

to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the' composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

I Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits-are provided to assure compatibility of operation with the fatigue analysis l

_ performed in accordance with the ASME Code requirements.

l

-The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350 F.

l Either PORV has adequate relieving capability to protect the RCS from overpres-surization when the transient is limited to either: (1) the start of an idle l

RCP with the secondary water temperature of the steam generator less than or l

equal to 50'F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water solid RCS.

These two scenarios are analyzed to determine the resulting overshoots assuming a single-PORV actuation with a stroke time _of 2.0 seconds from full l

closed to full open.

Figure 3.4-4a (3.4-4b) are based upon this analysis and I

represents the maximum allowable PORV variable setpoint such that, for the two overpressurization transients noted, the resulting pressure will not exceed the Appendix G reactor vessel NDT limits.

3/4.4.10 STRUCTURAL INTEGRITY l

The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness L

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 l

10'CFR 50.55a(g) except where specific written relief has been granted by the l

Commission pursuant to 10 CFR 50.55a(g)(6)(i).

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-16 AMENDMENT NO. 30 i

_ _. _ - _ ~ _. _, __ _ _. _ _..., _. _. _ _. _ _ _, _.. _,.... _.,. _ _ _,,.. _.. _. ~ _. _ _ - - -. _ _., _,.,.., -