Text

Amends 54 & 53 to Licenses DPR-80 & DPR-82,respectively, Revising Tech Specs to Change Heatup & Cooldown Curves & Delete Surveillance Withdrawal Schedule,Per Generic Ltr 88-11 & Rev 2 to Reg Guide 1.99
	ML20043A888
Person / Time
Site:	Diablo Canyon
Issue date:	05/11/1990
From:	Larkins J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20043A889	List: ML20043A888; ML20043A890;
References
	RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9005230243
	Download: ML20043A888 (23)
	v • d • e

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N' UNITED' STATES -

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- NUCLEAR REGULATORY COMMISSION '

al WASHINGTON, D. C. 20666

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PACIFIC GAS AND ELECTRIC COMPANY ~

i DIABLO CANYON NUCLEAR POWER PLANT, UNIT-1 1

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DOCKET NO.'50-275 1

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AMENDMENT TO FACILITY OPERATING LICENSEL

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Amendment No. 54 i

4 License-No.'DPR-80 j

1.'

The Nuclear Regulatory Comission-(the Comission)' has found -.that:'

A.

The application for' amendment by Pacific Gas &' Electric Companyt..

(the licensee), dated July 7,1989, complies with the; standards and-requirements.of the Atomic Energy Act of 1954,' as amended (the Act), and the Commission's regulations set'forth.in 10 CFR Chapter I,

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B.

-The facility will o)erate in conformity with~the application, J,

the provisions of: tie Act and the regulaticas of-the Comission;.

C.

There is reasonable assurance'(i)- that the: activities authorized l

by this amendment can be conducted without endangering the health t

g 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'

-i defense and security or to the' health and safety of the'public; i

and E.

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

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2 2 z _Accordingly, the-license is amended by changes to the-Technical Specifications as-indicated in the attachment to this> license amendment,1and paragraph 2.C.(2) of Facility Operating License i

No. DPR-80 is_hereby amended.to read as follows:'

(2) Technica1' Specifications The Technical Specifications. contained-in Appendix A and the' t

Environmental Protection Plan-contained in Appendix B, as revised through Amendment No. 54', are hereby; incorporated in the license..

Pacific Gas & Electric Company shall operate:the facility in accordance with the Technical Specifications and the Environmental.

Protection Plan, except where otherwise_ stated in specific license conditions.

3.

This license amendment becomes effective at the date of its issuance.-

FOR THE NUCLEAR REGULATORY COMMISSION' John T. Larkins, Acting: Director:

Project Directorate V Division of Reactor-Projects -_III, IV, Y and Special Projects-Office of Nuclear Reactor Regulation-1

Attachment:

Changes.to the Technical

= Specifications Date of Issuance:

May 11, 1990 I

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. UNITED STATES -

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NUCLEAR REGULATORY COMMISSION

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*. *,o PACIFIC GAS AND ELECTRIC COMPANY L

DIABLO CANYON' NUCLEAR POWER PLANT, UNIT 2-DOCKET NO2 50-323' AMENDMENT TO FACILITY. OPERATING LICENSE.

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Amendment No. 53

' License No. DPR 1

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. The Nuclear Regulatory Commission (the Commission)~ ha's found that:.

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A.

The application:for amendment by, Pacific Gas'a Electric Company (the licensee), ' dated July 7,1989, complies with the standards and '

i requirements'of the Atomic Energy Act of 1954, as: amended (the-Act), and the Commission's regulations set forth in.10 CFR-i Chapter I; B.-

The facility will o)erate in' conformity:with'the' application, the. provisions-of. tle Act, and the regulations of. the-

. Commission; C.

-There is reasonable assura'r.ce-(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, A'

L D.

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

The issuance of this amendment'is in:accordance with 10 CFR Part 51 D

of the Comission's. regulations and all applicable requirements have L

been satisfied.

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'Accordingly, the license is amended by' changes to the' Technical.

Specifications as indicated in the attachment to this license-

- l amendment,'and paragraph 2.C.(2) of Facility Operating. License i

No. DPR-82"is hereby amended to read as follows:.

(2). Technical Specifications

~The Technical Specifications contained in Appendix A and the1 Environmental' Protection. Plan contained -in = Appendix. B. as revised 4

.through' Amendment-No. 53, are hereby. incorporated.in the license.

Pacific Gas & Electric Company shall operate the facility 11n

.accordance with.the Technical. Specifications and the' Environmental' LProtection Plan, except where otherwise stated in; specific license conditions.

3.

This license' amendment becomes effective at.the date of.its issuance.-

FOR THE NUCLEAR REGULATORY COMMISSION l

.p John T. Larkins, Acting Director Project-Directorate Y

- Division of: Reactor Projects

.III, i

IV,.V and~Special Projects; Office of Nuclear Reactor Regulation;

Attachment:

Changes to the Technical i.

Specifications Date of Issuance: May 11, 1990 l

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ATTACliMENT TO LICENSE AMENDMENT NOS. 54 AND 53 FACILITY OPERATING LICENSE NOS. DPR-80 and'DPR-82

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DOCKET NOS. 50-275-AND 50-323 1

4 Replace-the following pages.of the~ Appendix "A" Technical Specifications'with the' attached'pages. The revised pages are identified by amendment number and t

contain vertical lines indicating the areas of change. -0verleaf pages.are~

t also included, as' appropriate..

Remove Page Insert Page-4 viii-viii l

Xv XV-

3/4 4-30 3/4 4,

3/4 4-31

'3/4-4-31

-3/4:4-32 3/4'4-32 3/4 4-33 3/4 4 B 3/4 4.B.3/4 4-7

-B 3/4 4-8 B 3/4 4 1 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 i

B.3/4 4-13 B 3/4 4-13~

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x INDEX-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-i SECTION.

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3/4.3 -INSTRUMENTATION (continued)-

l Chlorine ~ Detection Systems............................'...

3/4 3-54 Fire Detection Instrumentation..........................

3/4 3-55 g

- TABLE 3. 3-11 '. FI RE DETECTION ' INSTRUMENTS..........................

3/4 3-56 Radioactive' Liquid Effluent Monitoring Instrumentation..

3/4 3-59

. TABLE 3.3-12 RADI0 ACTIVE LIQUID _ EFFLUENT. MONITORING i

i INSTRUMENTATION..........................................

3/4 3 TABLE 4.3 -

t!

RAD 10 ACTIVE' LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-62 o

Radioactive Gaseous' Effluent Monitoring Instrumentation.

3/4 3-64

' TABLE-3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING H

INSTRUMENTATION.........................................

3/4 3-65 h

TABLE 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-67 i

3/4.3.4

' TURBINE OVERSPEED PROTECTION............................

3/4-3-69 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation............................'.

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

3/4 4-2' l

l Hot Shutdown............................................

3/4 4-3 Cold Shutdown - Loops F111ed............................

3/4 4-5 Cold Sliutdown - Loops Not F111ed.......................

3/4 4-6 i

3/4.4.2.

SAFETY VALVES Shutdown................................................

3/4 4-7 Operating...............................................

3/4 4-8 4

- 3/4.4.3 PRESSURIZER.............................................

3/4 4-9 3/4.4.4 RELIEF VALVES...........................................

3/4 4-10 3/4.4.5 STEAM GENERATORS........................................

3/4 4-11 DIABLO CANYON - UNITS 1 & 2 vii 1

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INDEX LIMITING CONDITIONS FOR'0PERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM (continued)

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE'

'I

. INSPECTED DURING INSERVICE INSPECTION..........'.........

3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION......................

3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...............................

'3/4 4-18 Operational Leakage...................................

-3/4 4 19 j

TABLE 3.4-1 REACTOR-COOLANT SYSTEM PRESSURE ISOLATION VALVES.....:'3/4 4-21 3/4.4.7 CHEMISTRY...............................................

.3/4 4-22 I

i TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS..............

3/4'4-23 TABLE 4.4-3' REA;IOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS...............................

3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY........................................

3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT'.1-131 REACTOR COOLANT SPECIFIC.

l ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

> 1.pCI/ GRAM DOSE EQUIVALENT I-131......................

3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE.

AND ANALYSIS PR0 GRAM....................................

3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS ~............................

3/4 4-30' FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 8 EFPY................................

'3/4 4-31

(

i FIGURE 3.4-3 REACTOR C00LANT' SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 8 EFPY................................

3/4 4-32 Pressurizer.........................................'....

3/4 4-34 Overpressure Protection Systems.........................

3/4 4-35 3/4.4.10 STRUCTURAL INTEGRITY....................................

3/4 4-37 3/4.4.11 REACTOR VESSEL HEAD VENTS...............................

3/4 4-38 DIABLO CANYON - UNITS 1 & 2 viii Amendment Nos.54 and 53 l

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INDEX J

. BASES J

SECTION PAGE s

. 3/4.4.6-REACTOR ~ COOLANT SYSTEMLEAKAGE...........................

B 3/4 4-3' l

U 3/4.4.7 -. CHEMISTRY................................................

B 3/4.4-5 i

3/4.4.B-SPECIFIC ACTIVITY.'.......................................

B 3/4.4-5 3

-3/4.-4.9. PRESSURE / TEMPERATURE LIMITS..................'............

B 3/4 4 :

-3/4.4.10 STRUCTURAL ~ INTEGRITY.....................................

,B 3/4:4-16 i!

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'3/4. 4.11 REACTOR. VESSEL HEAD VENTS.~................................

.B 3/4-4-16

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-3/4.5' EMERGENCY CORE COOLING SYSTEMS s

3/4.5.1' ACCUMULATORS.............................................

B 3/4 5-1

.e 3/4.5.2 and 3/4.5.3-ECCS SUBSYSTEMS...............'..............

~B 3/4 5-l' 5

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3/4.5.4-. BORON INJECTION SYSTEM...................................

lB 3/4 5-2 3/4.5.5 -REFUELING WATER STORAGE-TANK......................'.......

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1 DIABLO CANYON - UNITS 1 & 2 xv Amendment Nos54 and 53 i

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INDEX BASES i

i SECTION gGE r

3/4.6. CONTAINMENT SYSTEMS

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3/4.6.1 CONTAINMENT...............................s...............

6 ?/4 6-1

'3/4.6.2 DEPRESSURIZATION AND COOLING' SYSTEMS.....................

B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................

B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L..................................

B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................

B 3/4 7-1 3/4.7.2. STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........

B 3/4 7-3

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3/4.7.3 VITAL COMPONENT COOLING WATER SYSTEM.....................

B 3/4 7-3 i

3/4.7.4 AUXILIARY SALTWATER SYSTEM................................

t 3/4 7 3 3'/4.7.5 CONTROL ROOM VENTILATIONSYSTEM...........................

B 3/4 7-3 3/4.7.6 AUXILIARY BUILDING SAFEGUARDS AIR FILTRATION SiSTEN....

4 8 3/4 7-4 f

3/4.7.7 SNUBBERS.....................................

B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION.......................-.......

B 3/4 7-6 I

3/4.7.9 FIRE SUPPRESSION SYSTEMS..

3.............................

8 3/4 7-6 3/4.7.10 FIRE BARRIER PENETRATIONS................................

B 3/4 7-7 1

3/4.7.11 AREA TEMPERATURE MONITORING..............................

B'3/4 7-7 3/4.7.1?

'IMATE HEAT SINK.......................................

B 3/4 7-7 3/4.7.1?

J0D PROTECTION..............'...........................

B 3/4 7-8 3/4.8 Ett 'TRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and DNSITE 00WER DISTRIBUTION...............................

B 3/4 8-1

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3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.................'.

B 3/4 8-3 k

DIABLO CANYON - UNITS 1 &'2 xv1 i

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TABLE 4.4-4 (Continued) o-j TABLE NOTATIONS

ntil the specific activity of the Reactor Coolant System is restored within U

l its limits.

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Sample to be taken after 3 minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

- A gross radioactivity analysis shall consist of the quantitativs s:easurement of the total specific activity of the reactor coolant except for radionu-c11 des with half-lives less than 10 minutes and all radiciodines.

The total specific activity shall be the sum of the degassed beta gamma activity and i

the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample is taken and extrapolated back to when the sample was taken.

i Determination of the. contributors to the gross specific activity shall be l

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

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

a

- - A radiochemical analysis for I 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 radiciodines, which is identified in the reactor coolant.

The specific activities for these individual radionu-i clides shall be used in the determination of I for the reactor coolant

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sample.

Determination of the contributors to I shall be based upon those

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energy peaks identifiable with a 95% confidence level.

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'l DIABLO CANYON. UNITS 1 & 2 3/4 4-29 l

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_ REACTOR C00'LANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 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 3.4-2 and 3.4-3 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.

A maximum temperature change of,?ess than or equal to 10'F in any c.

1-hour period during inservice hydrostatic and leak testing operations above tie 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 withir. 30 minutes}* perform an engineering evaluation to determine the effects of the out-of-1 mit 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1 The Reactor Coolant System temperature and pressure shall be deter-mined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

DIABLO CANYON - UNITS 1 & 2 3/4 4-30 Amendment Nos. 54 and 53

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REACTOR C00LMT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP'TO 8 EFPY DIABLO CANYON - UNITS 1 & 2 3/4 4-31 Amendment Nos. 54 and 53 s

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REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 8 EFPY j

DIABLO CANYON - UNITS 1 & 2 3/4 4-32 Amendment Nos. 54 and 53 I

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REACTOR COOLANT SYSTEM l

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PRES $URIZER i

LIMITING CONDITION FOR OPERATION I

l 3.4.9.2 The pressurizer. temperature shall be limited to:

I s.

A maximum heatup of 100'F in any 1-hour period, i

b.

A maximum cooldown of 200'F in any 1-hour period, and i

c.

A maximum spray water temperature differential of 560'F.

I APPLICABILITY:

At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition

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on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY l

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less=than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVEILLANCE REQUIREMENTS l

L 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per hour during system heitup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

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REACTOR COOLANT SYSTEM BASES 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 4

Presure Vessel Code, Section !!!, Appendix G:

1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the excep' tion of the pressurizer) shall be Ifmited in accordance 4

with Figures 3.4-2 and 3.4-3 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 between those presented may be obtained by interpolation; and b.

Figures 3.4-2 and 3.4-3 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 l

i 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 generator 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 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 560'F, and i

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.

Allowable pressures and temperatures for inservice leak and hydrostatic tests are given in Figure 3.4-2.

6.

The criticality limit on Figure 3.4-2 is based on the minimum allowable temperature of 295'F for an inservice hydrostatic test of 110% of operating pressure.

The fracture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1966 Edition fo~ Unit 1 and the r

1968 Edition for Unit 2 of the ASME Boiler and Pressure Vessel Code, Sec-tion III. These properties are then evaluated in accordance with the NRC Standard Review Plan.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RTNDT, at the end of 8 effec-l tive full power years (EFPY) of service life. The 8 EFPY service life period l

6 is chosen such that the limiting RT ct the 1/4T location in the core region NDT DIABLO CANYON - UNITS 1 & 2 B 3/4 4-7 Amendment Nos. 54 and 53 w

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DIABLO CANYON - UNITS 1 & 2 B 3/4 4-8 Amendment Nos'. 54 and 53 Y

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LDIABLO CANYON - UNITS 1 & 2 B 3/4 4-11

' Amendment Nos. 54 and 53 t

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REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) is greater than the RT of the limiting unirradiated material. The selection NDT of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials h&ve been tested to determine their initial RTNDT; the results of these tests are shown in the FSAR Update.

Reactor oper-ation and resultant fas't neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, based NDT.

i upon the fluence, copper content and nickel content of the material in question, can be predicted using value of ARTNDT computed by Regulatory Guide 1.99, Rev-1sion 2 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," for the maximum neutron fluence at the locations of interest.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include pre-dicted adjustments for this shift in RTNDT at the end of 8 EFPY.

-l Values of ART determined in this manner will be used until the results l

NDT from the material surveillance program,in accordance with the requirement evaluated according to ASTM E185-82, can be used.

Capsules will be removed ASTM E185 and 10 CFR Part 50, Appendix H.

The surveillance specimen with-drawal schedule will be maintained in the FSAR Update.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel.

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

NDT Allowable pressure-temperature relationships for various heatup and cool-down rates are calculated using methods derived from Appendix G in Section III 4

of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR 1

Part 50 and these methods are discussed in detail in the following paragraphs.

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 i

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 3

operation 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 i

DIABLO CANYON - UNITS 1 & 2 B 3/4 4-12 Amendment Nos. 54 and 53 i.

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

REACTOR COOLANT $YSTEM BASES PRES $URE/ TEMPERATURE LIMITS (Continued) 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 3

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

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

The K curve is given by the equation:

gg Kgg = 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 gg 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:

CKgg + Kgg $, Kgg (2)

Where, Kgg = the stress intensity factor caused by membrane (pressure) stress, kit = i.he dress intensity factor caused by the thermal gradients, KIR = function of temperature relative to the RTreference stress intensity factor provid NDT of the material,-

1 C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations, i

At any time during the heatup or cooldown transient, K is determined IR by the metal temperature of 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 cal-culated and then the corresponding thermal stress intensity factors, Kgg, for the reference flaw are computed.

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

DIABLO CANYON - UNITS 1 & 2 B 3/4 4-13 Amendment Nos. 54 and 53

REACTOR COOLANT SYSTEM BASES i

PRESSURE / TEMPERATURE LIMITS -(Continued) f 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.

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

From wese 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, During cooldown, the 1/4T vessel location is at-a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situa-tion.

It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T location IR for finite cooldown r6tes than for steady-state operation.

Furthermore, if conditions exist such that the increase in K exceeds K3g, the calculated IR 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 1

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 ano assures conservative operation of the system for the entire cooldown period.

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 4

inside of the vessel wall.

The thermal gradients during heatup produce com-pressive stresses at the inside of the wall that alleviate the tensile stresses produced by_ internal pressure. The metal temperature at the crack tip legs the coolant temperature; therefore, the K for the 1/4T crack during heatup is IR lower than the K for the 1/4T crack during steady-state conditions at the yp i

same coolant temperature.

During heatup, especially at the end of the tran-

.sient, conditions may exist such that the effects of compressive thermal stresses and different K

's for steady-state and finite heatup rates do not IR of(set each other and the pressure-temperature curve based on steady-state l

conditions no longer repress.its a lower bound of all similar curves for finite g

DIABLO CANYON - UNITS 1 & 2 B 3/4 4-14 f

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ML20043A888

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