Text

Amends 37 & 36 to Licenses DPR-80 & DPR-82,respectively, Revising Tech Specs to Permit Core Reload & Operation W/ Westinghouse Fuel Assemblies
	ML20247B110
Person / Time
Site:	Diablo Canyon
Issue date:	05/10/1989
From:	Knighton G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20247B115	List: ML20247B110; ML20247B120;
References
	NUDOCS 8905240010
	Download: ML20247B110 (62)
	v • d • e

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UNITED STATES NUCLEAR REGULATORY COMMISSION 3

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PACIFIC GAS AND ELECTRIC COMPANY l

DIABLO CANYON NUCLEAR POWER PLANT, UNIT I DOCKET NO. 50-275 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-80 1.

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

A.

The application for amendment by Pacific Gas & Electric Company (thelicensee),datedNovember 29, 1988, as supplemented by letters dated December 9,1988 and February 17, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 regulations of the Commission; C.

Thereisreasonableassurance(i)thattheactivitiesauthorized 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.

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2.

Accordir,1y, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 37, 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 upon completion of Cycle 3 for Unit 1.

FOR THE NUCLEAR REGULATORY COMMISSION yN Mj' f%'

Georger. Knighton / irector Projec Directorate V Division of Reactor Projects - III, IV. Y and Special Projects Di*ic.; of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 10, 1989 1

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UNITED STATES 8 } h }t NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 h;.v /

PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 DOCKET NO. 50-323 AMENDMENT TO FACILITY OPERATING L,ICENSE Amendment No. 36 License No. DPR-82 1.

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

A.

The application for amendment by Pacific Gas & Electric Company (the licensee), dated November 29, 1988, as supplemented by letters dated December 9, 1988 and February 17, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 regulations of the Comission; 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:

-(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 36, 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 upon completion of Cycle 3 for Unit 1.

FOR THE NUCLEAR REGUL TORY COMMISSION sif f66 George

Knighton, 1 rector Project Directorate V Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 10,1989

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ATTACHMENT TO LICENSE AMENDMENT NOS. 37 AND 36 FACILITY OPERATING LICENSE NOS. DPR-80 and DPR-82 DOCKET NOS. 50-275 AND 50-323 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 areas of change.

Overleaf pages are included, as app 7priate.

Remove Page Insert Page iii iii y

y xxii xxii 2-1 2-1 2-2 2-2 2-2a 2-2u 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 B 2-1 B 2-1 B 2-la B 2-lb B 2-4 B 2-4 3/4 1-20 3/4 1-20 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-5 3/4 2-5 3/4 2-7 through 3/4 2-17 3/4 2-7 through 3/4 2-22 3/4 3-23 3/4 3-23 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 10-2 3/4 10-2 B 3/4 2-4 8 3/4 2-4 8 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 6-20 6-20

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I INDEX

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION Page 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.....................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..................................

2-1 FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - UNITS 1&2 Cycle 4 and After...

Er2 FIGURE 2.1-1b REACTOR CORE SAFETY LIMIT - UNIT 2 Cycle 3................

2-2a 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP01NTS....................

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0INTS.........

2-4 BASES SECTION Page 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.....................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..................................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS.....................

B 2-3

.DIABLO CANYON - UNITS 1 & 2 fii Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3

s INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

' SECTION pg 3/4.0 APPLICABILITY..............................................

3/4 0-1

'3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg greater than 200*F...............

3/4 1-1 Shutdown Margin - T,yg less than or equal to 200*F......

3/4 1-3' Moderator Temperature Coefficient.......................

3/4 1-4 Minimum Temperature for Criticality.....................

3/4 1-6

- 3/4.1.2-BORATION SYSTEMS Flow Path - Shutdown....................................

3/4 1-7 Flow Paths - Operating..................................

3/4 1-8 Charging Pump - Shutdown................................

.3/4 1-10 1

Charging Pumps - Operating..............................

3/4 1-11 Borated Water Source - Shutdown.........................

3/4 1-12 Borated Water Sources -

Operating.......................

3/4 1-13 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................

3/4 1-15 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT 0F AN INOPERABLE FULL-LENGTH R00..................

3/4 1-17 Position Indication Systems - Operating.................

3/4 1-18 Position Indication System - Shutdown...................

3/4 1-19 Rod Drop Time...........................................

3/4 1-20 Shutdown Rod Insertion Limit............................

3/4 1-21

. Control Rod Insertion Limits........

3/4 1-22 DIABLO CANYON - UNITS 1 & 2 iv

f f

INDEX

~

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Page 3/4.1 REACTIVITY CONTROL SYSTEMS (Continued) i FIGURE 3.1-la R0D BANK INSERTION LIMITS VERSUS THERMAL POWER (UNIT 1).........................................

3/4 1-23 FIGURE 3.1-lb R0D BANK INSERTION LIMITS VERSUS THERMAL POWER (UNIT 2).........................................

3/4 1-24 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................

3/4 2-1 FIGURE 3.2-la AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (Units 1 and 2 Cycle 4 and After)............................................

.3/4 2-2 FIGURE 3.2-lb AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (Unit 2 Cycle 3)......

3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F (Z).....................

3/4 2-5 9

FIGURE 3.2-2 K(Z) - NORMALIZED F HEIGHT............ 0.(.Z) AS A FUNCTION OF CORE 3/4 2-6 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.....................................

3/4 2-13 FIGURE 3.2-3a RCS TOTAL FLOWRATE VERSUS R (UNIT 1 Cycle 4 and After)............................................

3/4 2-14 FIGURE 3.2-3b RCS TOTAL FLOWRATE VERSUS R (UNIT 2 Cycle 3)......

3/4 2-15 FIGURE 3.2-3c RCS TOTAL FLOWRATE VERSUS R (UNIT 2 Cycle 4 and After)............................................

3/4 2-16 3/4.2.4 QUADRANT POWER TILT RATI0..............................

3/4 2-18 3/4.2.5 DNB PARAMETERS.........................................

3/4 2-21 TABLE 3.2-1 DNB PARAMETERS......................................

3/4 2-22 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................

3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................

3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES..

3/4 3-8 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUME"Te r10N SURVEILLANCE REQUIREMENTS................

3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................

3/4 3-14 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-15 DIABLO CANYON - UNITS 1 & 2 v

Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3

f I

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Pg

]

l 3/4.3 INSTRUMENTATION (continued)

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS..........................

3/4 3-23 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............

3/4 3-28 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-32 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................

3/4 3-36 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..............................................

3/4 3-37 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.....................

3/4 3-39 Movable Incore Detectors.................................

3/4 3-40 Seismic Instrumentation..................................

3/4 3-41 TABLE 3.5-7 SEISMIC MONITORING INSTRUMENTATION....................

3/4 3-42 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................

3/4 3-43 Meteorological Instrumentation...........................

3/4 3-44 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.............

3/4 3-45 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-46 Remote Shutdown Instrumentation..........................

3/4 3-47 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............

3/4 3-48 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................

3/4 3-49 Accident Monitoring Instrumentation......................

3/4 3-50 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..

3/4 3-52 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................

3/4 3-53 DIABLO CANYON - UNITS 1 & 2 vi

4' INDEX

' ADMINISTRATIVE CONTROLS SECTION PAGE 6.5 REVIEW AND AUDIT (Coniinued) 6.5.2 GENERAL OFFICE NUCLEAR PLANT REVIEW AND AUDIT COMMITTEE (GONPRAC)

Function........................................................

6-9 Composition....................................................

6-10 Alternates.....................................................

6-10 Consultants....................................................

6 Meeting Frequency..............................................

6-10 Quorum.........................................................

6-10 Review.........................................................

6-10 Audits.........................................................

6-11 Records........................................................

6-12 6.6. REPORTABLE EVENT ACTI0N..........................................

6-12 6.7 SAFETY LIMIT VIOLATION............................................

6-13 6.8 PROCEDURES AND PR0 GRAMS..........................................

6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS................................................

6-16 Startup Reports................................................

6-16 j

Annual Reports.................................................

6-16 Annual Radiological Environmental Operating Report.............

6-17 d

Semiannual Radioactive Effluent Release Report.................

6-18 Monthly Operating Report.......................................

6-20 Peaki ng Factor Limi t Report....................................

6-20

' 6.9.2 SPECIAL REP 0RTS................................................

6-20 6.10 RECORD RETENTION................................................

6-20 6.11 RADIATION PROTECTION PR0 GRAM....................................

6-22 6.12 HIGH RADIATION AREA.............................................

6-22 l

DIABLO CANYON - UNITS 1 & 2 xxii Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3

\\

5 INDEX LIMITING CONDITIONS FOR OPERATION ~AND~SURVEIEEANCE REQUIREMENTS SECTION g

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................

3/4 11-1 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM..................................................

3/4 11-2 Dose...................

3/4 11-5 Liquid Radwaste Treatment System.........................

3/4 11-6 Liquid Holdup Tanks......................................

3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................

3/4 11-8 TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM..................................................

3/4 11-9 Dose'- Noble Gases.......................................

3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form.................

3/4 11-13 Gaseous Radwaste Treatment System........................

3/4 11-14 Explosive Gas Mixture....................................

3/4 11-15 Gas Storage Tanks........................................

3/4 11-16 3/4 11.3 SOLID RADIOACTIVE WASTE..................................

3/4 11-17 3/4 11.4 TOTAL 00SE...............................................

3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING

'3/4.12.1 MONITORING PR0 GRAM.......................................

3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........

3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES.................................

3/4 12-8 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS.................................................

3/4 12-9 3/4.12.2 LAND USE CENSUS..........................................

3/4 12-12 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM.......................

3/4 12-13 DIABLO CANYON - UNITS 1 & 2 xiii

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-la for Units 1 and 2 Cycle 4 and after, and Figure 2.1-lb for Unit 2 Cycle 3.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer

-pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.7.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.

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DIABLO CANYON - UNITS 1 & 2 2-1 Amendment Nos.37 and 36 q

Effective at end of Unit 1 Cycle 3 s

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DIABLO CANYON - UNITS 1 & 2 2-2 Amendment Nos.37 and 36 Effective at end of Unit 1 Cycle 3 J

UNIT 2 CYCLE 3 680 870 UNACCEPTABLE' l

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REACTOR CORE SAFETY LIMIT (UNIT 2 CYCLE 3) i DIABLO CANYON - UNITS 1 & 2 2-2a Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3

' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Trip System Instrumentation or Interlock Setpoint less con-servative than the value shown in the Allowable Values column of Table.2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE.

status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

DIABLO CANYON - UNITS 1 & 2 2-3

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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Units 1 and 2 Cycle 4 and after The DNB design basis is as follows:
there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during steady-state opera-tion, normal operational transients, and anticipated transients is greater than or equal.to the DNBR limit of the DNB correlation being used (the WRB-1 for LOPAR fuel and the WRB-2 for VANTAGE 5 fuel in this application).
The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with a 95 percent confidence level that DNB will not occur when the minimum DNBR is at or greater than the DNBR limit (1.17 for both the WRB-1 and WRB-2 correlations).
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability with a 95 confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit.
The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty.
This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertain-ties.
For Diablo Canyon Units, the design DNBR values are 1.33 and 1.37 for thimble and typical cells, respectively, for LOPAR fuel, and 1.30 for thimble I
and 1.32 for typical cells for the VANTAGE 5 fuel.
In addition, margin has been maintained in both designs by meeting safety analysis DNBR limits of 1.44 for thimble and 1.48 for typical cells for LOPAR fuel, and 1.68 and 1.71 for thimble and typical cells, respectively, for VANTAGE 5 fuel in performing safety analyses.
The curves in Figure 2.1-la show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the safety analysis DNBR limits, or the average enthalpy
)
at the vessel exit is less than the enthalpy of saturated liquid.
The curves are based on an enthalpy hot channel factor, F f 1.56 for LOPAR H
and 1.59 for VANTAGE 5 fuel, and a reference cosine with a peak of 1.55 for axial power shape.
An allowance is included for an increase in F at reduced H
power based on the expressions:
DIABLO CANYON - UNITS 1 & 2 B 2-1 Amendment Nos.37and 36 Effective at end of Unit 1 Cycle 3
r 2.1 SAFETY LIMITS BASES (Continued)
F g = 1.56 [1 + 0.3(1 - P)] for LOPAR fuel Ffg=1.59[1+0.3(1-P)] for VANTAGE 5 fuel where P is the fraction of RATED THERMAL POWER The4%measurementuncertaintyassociatedwithFfH is accounted for in the DNBR design limit.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip.
When the axial power imbalance 2is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trip will reduce the Setpoints to provide protection consistent with core Safety Limits.
Unit 2 Cycle _3 Operation above the upper boundary of the nucleate boiling regime could result in excessive claddin from nucleate boiling (DNB)g temperatures because of the onset of departure and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the R-Grid correlation.
The R-Grid DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.
The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curves of Figure 2.1-1b show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
The curves are based on an enthalpy hot channel factor, F f 1.55 and H
a reference cosine with a peak of 1.55 for axial power shape.
An allowance is included for an increase in FfH at reduced power based on the expression:
F H = 1.55 [1+ 0.3 (1-P)]
where P is the fraction of RATED THERMAL POWER The4%measurementuncertaintyisincludedintheFfH value.
DIABLO CANYON - UNITS 1 & 2 B 2-la Amendment Nos.37and 36 Effective at end of Unit 1 Cycle 3
l 2.1 SAFETY LIMITS BASES (Continued)
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip.
When the axial power imbalance 2
-is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trip will reduce the Setpoints to provide protection consistent with core Safety Limits.
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1 DIABLO CANYON - UNITS 1 & 2 B 2-lb Amendment Nos.37and 36 Effective at end of Unit 1 Cycle 3 I
I 2.2 LIMITING SAFETY SYSTEM SETTINGS 4
BASES 2.2.1 REACTOR TRI) SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.
In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing protection system fu'nctional
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diversity.
1 The Reactor Trip' System initiates a Turbine trip signal whenever Reactor trip is initiated.
This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldewn and thus avoids unneces-sary actuation of the Engineered Safety Features Actuation System.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance for all trips including those trips assumed in the safety analyses.
Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.
Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting.
The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
DIABLO CANYON - UNITS 1 & 2 B 2-3
LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued)
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 setpoint.
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.
The Power Range Negative Rate Trip provides protection for control rod drop' accidents.
At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.
The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor.
No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBRs will be greater than or equal to the DNBR limits.
Intermediate and Source Range, Neutron Flux The Intermediate and Source Range Neutron Flux trips provide core protec-tion during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.
These trips provide redundant protection to the Low Setpoint trip of the Power Range Neutron Flux channels.
The Source Range channels will initiate a Reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes active.
The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
No credit was taken for operation of the trips asso-ciated with either the Intermediate or Source Range channels in the accident analyses; however, tneir functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System.
l Overpower AT The Overpower AT trip provides assurance of fuel integrity, e.g., no fuel pellet cracking or melting, under all possible overpower conditions, limits the required range for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied i
l DIABLO CANYON - UNITS 1 & 2 B 2-4 Amendment Nos.37and 36 l
Effective at end of Unit 1 Cycle 3
I REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding demand position indi-cation) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted.
APPLICABILITY:
MODES 3*#, 4*# and 5*#.
~
ACTION:
With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.
SURVEILLANCE REQUIREMENTS I
4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full range of rod travel at least once per 18 months, i
With the Reactor Trip System breakers in the closed position.
See Special Test Exceptions Specification 3.10.4 DIABLO CANYON - UNITS 1 & 2 3/4 1-19
REACTIVITY CONTROL SYSTEMS R0D DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod, drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds for Units 1 and 2 Cycle.4 and after, and 2.2 seconds for Unit 2 Cycle 3 from beginning of decay of stationary gripper coil voltage to das' Jot entry with:
T,yg greater than or equal to 541 F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY:
MODES 1 and 2.
ACTION:
With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
a.
-For all rods following each removal of the reactor vessel head, b.
For specifically affected individual rods following any main-tenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.
At least once per 18 months.
DIABLO CANYON - UNITS 1 & 2 3/4 1-20 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITIONS FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-la for Units 1 and 2 Cycle 4 and after, and Figure 3.2-1b for Unit 2 Cycle 3.
APPLICABILITY:
MODE 1 AB0VE 50 PERCENT RATED THERMAL POWER *.
ACTION:
a.
With the indicated AXIAL FLUX DIFFERENCE outside of the Figure 3.2-la limits or Figure 3.2-1b limits, 1.
Either restore the indicated AFD to within the Figure 3.2-la or Figure 3.2-1b limits within 15 minutes, or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
b.
THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-la or Figure 3.2-1b limits.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50 percent of RATED THERMAL POWER by:
a.
Monitoring the indicated AFD for each OPERABLE excore channel:
1.
At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2
At least once per hour when the AFD Monitor Alarm is inoperable.
4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits.
l 1
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j OSee Special Test Exceptions Specification 3.10.2 DIABLO CANYON - UNITS 1 & 2 3/4 2-1 Amendment Nos. 37 and 36 I
Effective at end of Unit 1 Cycle 3 i
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FIGURE 3.2-la AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (UNITS 1 AND 2 CYCLE 4 AND AFTER)
DIABLO CANYON - UNITS 1 & 2 3/4 2-2 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
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FIGURE 3.2-lb AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (UNIT 2 CYCLE 3) l DIABLO CANYON - UNITS 1 & 2 3/4 2-3 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
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i DIABLO CANYON - UNITS 1 & 2 3/4 2-4 Amendment Nos.12 and 10 i
POWER DISTRIBUTION LIMITS 3/4.2.2HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:
q F (Z) C [2.45] [K(Z)] for P > 0.5 0
P
~
F (Z) 1 [4.90] [K(Z)] for P 1 0.5 9
THERMAL POWER
, and Where P =
RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY:
MODE 1 (Units 1 and 2 Cycle 4 and after).
ACTION:
With F (2) exceeding its limit:
9 a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 9
within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit.
The Overpower 9
AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.
b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) 9 is demonstrated through incore mapping to be within its limit.
l DIABLO CANYON - UNITS 1 & 2 3/4 2-5 Amendment Nos. 37 and 36 l
Effective at end of Unit 1 Cycle 3 1
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CORE HEIGHT (FT)-
FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT q
DIABLO CANYON - UNITS 1 & 2 3/4 2-6 Amendment Hos.14 and 13
(
t POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.1.2 F (z) shall be evaluated to determine if F (z) is within its q
q limits by:
a.
Using the moveable incore detectors to obtain a power distribu-tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
b.
Increasing the measured F (z) component of the power distribution q
map by 3% to account for manufacturing tolerances and further in-creasing the value by 5% to account for measurement uncertainties, c.
Satisfying the following relationship:
F (z) 1 2.45 x K(z) for P > 0.5 P x W(z)
F (z) 5 2.45 x K(z) for P 1 0.5 W(z) x 0.5 where F (z) is the measured F (z) increased by the allowances q
for manufacturing tolerances and measurement uncertainty, 2.45 is the F limit, K(z) is given in Figure 3.2-2, P is the relative q
THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.
This function is given in the Peaking Factor l
Limit Report as per Specification 6.9.1.8.
d.
Measuring F (z) according to the following schedule:
1.
Upon achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q
2.
At least once per 31 Effective Full Power Days (EFPD),
whichever occurs first.
e.
With measurements indicating IM(z)\\
maximum F
over z ( K(z)j has increased since the previous determination of F (z) either of the following actions shall be taken:
During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
DIABLO CANYON - UNITS 1 & 2 3/4 2-7 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS I
SURVEILLANCE REQUIREMENTS (Continued) l L
1)
'F (z) shall be increased by 2% over that specified in Specification 4.2.2.1.2.c, or
~
2)
F (z) shall be measured at least once per 7 EFPD until two successive maps indicate that maximum 4(z)}isnotincreasing.
.over z I z/
f.
With the relationship specified in Specification 4.2.2.1.2.c above not being satisfied:
1)
Calculate the percent F (z) exceeds it limit by the following g
expression:
[ maximum F (z) x W(z)
-1 x 100 for P > 0.5 2.45 x K(z)
(overz
)-1)x100forP<0.5
~
[ maximum F (z) x W(z)
(overz' 2.45 x K(z)
/
J
_GT 2.
Either one of the'following actions shall be taken:
a)
Place.the core in an equilibrium condition where the limit in Specification 4.2.2.1.2.c is satisfied.
Power level may then be increased provided the AFD limits of Figure 3.2-la are reduced 1% AFD for each percent F (z) exceeds its limit, or 9
j 1
l
'l DIABLO CANYON - UNITS 1 & 2 3/4 2-8 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
P-POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREP.ENTS (Continued) b)
Comply with the requirements of Specification 3.2.2.1 for F (z) exceeding its limit by the percent calculated.
j q
g.
The' limits specified in Specification 4.2.2.1.2.c, 4.2.2.1.2.e,'and 4.2.,2.1.2.f above are not applicable in the following core plane regions:
1.
Lower core region from 0 to 15%, inclusive.
2.
Upper core region from 85 to 100%, inclusive.
4.2.2.1.3 When F (Z) is measured pursuant to Specification 4.10.2.2, an overall q
measured F (Z) shall be obtained from power distribution map and increased by q
3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
DIABLO CANYON - UNITS 1 & 2 3/4 2-9 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER-DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION i
i 3.2.2.2 F (Z) shall.be limited by the following relationships:
9 F (Z)'_<
2.45 [K(Z)] for P > 0.5 g
7 F (Z) 1 [4.64] [K(Z)] for P s 0.5 q
THERMAL POWER
, and Where P =
RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY:
MODE 1 (Unit 2 Cycle 3).
ACTION:
With F (Z) exceeding its limit:
q a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 9
within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION
'may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit.
The 9
Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least H0T STANDBY.
b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.
9 DIABLO CANYON - UNITS 1 & 2 3/4 2-10 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION 11MITS -
SURVEILLANCE REQUIREMENTS 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2.2 F
shall be evaluated to determine if F (Z) is within its limit by:
xy 9
a.
Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b.
Increasing the measured F component of the power distribution xy map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, c.
Comparing the F computed (Fx) btained in Specification xy 4.2.2. 2.2b., above, to:
1.
The F limits for RATED THERMAL POWER (FRTP) for the appro-xy x
priate measured core planes given in Specification 4.2.2.2.2e.
and f. below, and 2.
The relationship:
RTP F
= (Fxy ) [1+0.2(1-P)],
xy L
where F*Y is the limit for fractional THERMAL POWER operation RTP expressed as a function of F and P is the fraction of xy RATED THERMAL POWER at which F was measured.
xy d.
Remeasuring F according to the following schedule:
xy RTP 1.
When F is greater than the F limit for the appropriate x
measured core plane but less than the F relationship, addi-x tionalpowerdistributionmapsshallbetakenandFfcompared x
RTP to F and F either:
x x
a)
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL C
POWER or greater, the THERMAL POWER at which F*Y was last determined, or b)
At least once per 31 EFPD, whichever occurs first.
DIABLO CANYON - UNITS 1 & 2 3/4 2-11 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
.. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 1 2.
When the F is less than or equal to the (FRTP). limit for x
x the appropriate measured core plane, additional power RTP.
distribution maps shall be taken and F compared to F x
x and F at least once per 31 EFPD.
x e.
The F limit for RATED THERMAL POWER (FRTP) shall be provided xy x
for all core planes containing' Bank "D" control rods and all unrodded core planes in a Peaking Factor Limit Report per Specification 6.9.1.8.
f.
The F limits of Specification 4.2.2.2.2e., above, are not xy applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:
1.
Lower core region from 0 to 15%, inclusive, 2.
Upper core region from 85 to 100% inclusive, 3.
Grid plane regions at 17.8 t 2%, 32.1 1 2%, 46.4 i 2%,
60.6 i 2% and 74.9 1 2%, inclusive, and 4.
Core plane regions within i 2% of core height ( 2.88 inches) about the bank demand position of the Bank "D" control rods.
C l
g.
With F exceeding F
, the effects of F on F ifFQ(Z)iswithifiits19m(2)shallbe Y
its.
evaluat d to determin 4.2.2.2.3 When F (Z) is measured pursuant to Specification 4.10.2.2, an overall 9
measured F (Z) shall be obtained from power distribution map and increased by q
3%.to account for manufacturing tolerances and further increased by 5% to I
account for measurement uncertainty.
DIABLO CANYON - UNITS 1 & 2 3/4 2-12 Amendment Nos.37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3a for Unit 1 Cycle 4 and after, Figure 3.2-3b for Unit 2 Cycle 3, and Figure 3.2-3c for Unit 2 Cycle 4 and after for four loop operation.
Where:
N F
AH for LOPAR fuel (Units 1 and 2 Cycle 4 R
1.56 [1.0 + 0.3 (1.0 -P)] and after) a.
N F
~
AH For VANTAGE 5 fuel (Units 1 and 2 R
1.59 [1.0 + 0.3 (1.0 -P)] Cycle 4 and after)
N F
AH (Unit 2 Cycle 3)
R _ 1.49 [1.0 + 0.3 (1.0 -P)]
THERMAL POWER
, and b*
P _ RATED THERMAL POWER H=MeasuredvaluesofFfH btained by using the movable incore c.
F detectors to obtain a power distribution map.
The measured values of F shall be used to calculate R since Figure 3.2-3a H
for Unit 1 Cycle 4 and after includes measurement uncertainties of 2.4% for flow, Figure 3.2-3b for Unit 2 Cycle 3 includes measurement uncertainties of 3.5% for flow, and Figure 3.2-3c for Unit 2 Cycle 4 and after includes measurement uncertainties of 2.4% for flow.
All three Figures account fog measurement uncertainties of 4% for incore measurement of FAH' APPLICABILITY: MODE 1.
ACTION:
With the combination of RCS total flow rate and R outside the region of accept-able operation shown on Figure 3.2-3a for Unit 1 Cycle 4 and after, Figure 3.2-3b for Unit 2 Cycle 3 and Figure 3.2-3c for Unit 2 Cycle 4 and after:
a.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:
1.
Restore the combination of RCS total flow rate and R to within the above limits, or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
DIABLO CANYON - UNITS 1 & 2 3/4 2-13 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
I 48 MEASUREMENT 3
UNCERTAINTIES OF g
b 2.4% FOR FLOV AND l 46 o.
zi 4% FOR INCORE i
MEASUREMENT OF Ffs !
a
<r ARE INCLUDED IN O
a THIS FIGURE I
W I
ACCEPTABLE UNACCEPTABLE y
DPERATION OPERATION 0
40
[
REGION REGIDN L.
H 38 o
i-j 36 35.92) a:
0.90 0.95 1.00 1.05 1.10 1.15 i
R = F Es /1.56[1.0+0.3(1.0-P)] FOR LDPAR FUEL R = F Eu /1.59C1.0+0.3(1.0-P)] FOR VANT AGE 5 FUEL FIGURE 3.2-3a RCS TOTAL FLOWRATE VERSUS R (UNIT 1 CYCLE 4 AND AFTER)
DIABLO CANYON - UNITS 1 & 2 3/4 2-14 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
40 I
MEASUREMENT UNCERTAINTIES OF g
b 3.5% FOR FLOW AND 46 o_
5 4% FOR INCORE 1
MEASUREMENT OF Ffd 1
ARE INCL ~UDED IN 44 O
S THIS FIGURE J
w V
E ACCEPTABLE UNACCEPTABLE y
DPERATIDN DPERATION O
40 REGIDN REGION L.
N H
38 3
36
L 36.6) e 34 0.90 0.95 1.00 1.05 1.10 1.15 R = F [s[.49C1.0 + 0.3(1.0-P)]
~
FIGURE 3.2-3b RCS TOTAL FLOWRATE VERSUS R (UNIT 2 CYCLE 3)
DIABLO CANYON - UNITS 1 & 2 3/4 2-15 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
48 MEASUREMENT UNCERTAINTIES OF q
C-2.4% FDR FLOW AND o_
E 4% FOR INCORE U
-J L
MEASUREMENT OF FE'
<r ARE INCLUDED IN 4 '4 C
S
[
THIS FIGURE l
u W
h i
ACCEPTABLE UNACCEPTABLE l
. y DPERATION DPERATION I
3.
40 REGION REGION L.
__J 6
38 aw 36 ci, 36.25) a:
34 0.90 0.95 1.00 1.05 1.10 1.15 R = F Es /1.56[1.0+0.3(1.0-P)] FOR LDPAR FUEL R = F Es /1.59C1.0+0.3(1.0-P)] FOR VANT AGE 5 FUEL i
FIGURE 3.2-3c i
RCS TOTAL FLOW RATE VERSUS R (UNIT 2 CYCLE 4 AND AFTER)
DIABLO CANYON - UNITS 1 & 2 3/4 2-16 Amendment Nos. 37 and 36 l
Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) b.
Within 24 hourc of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
~
c.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b., above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3a for Unit 1 Cycle 4 and after, Figure 3.2-3b for Unit 2 Cycle 3, and Figure 3.2-3c for Unit 2 Cycle 4 and after, prior to exceeding the following THERMAL POWER levels:
1.
A nominal 50% of RATED THERMAL POWER, 2.
A nominal 75% of RATED THERMAL POWER, and 3.
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.2 The combination of indicated RCS total flow rate and R shall be deter-mined to be within the region of acceptable operation of Figure 3.2-3a for Unit 1 Cycle 4 and after, Figure 3.2-3b for Unit 2 Cycle 3, and Figure 3.2-3c for Unit 2 Cycle 4 and after:
a.
Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b.
At least once per 31 Effective Full Power Days.
4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3a for Unit 1 Cycle 4 and after, Figure 3.2-3b for Unit 2 Cycle 3, and Figure 3.2-3c for Unit 2 Cycle 4 and after at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the value of R, obtained per Specification 4.2.3.2, is assumed to exist.
4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
4.2.3.5 The RCS total flow rate shall be determined by measurement at least once per 18 months.
l DIABLO CANYON - UNITS 1 & 2 3/4 2-17 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO L
LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY:
MODE 1 ABOVE SD% OF RATED THERMAL POWER *.
ACTION:
a.
With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
1.
Calculate the QUADRANT POWER TILT RATIO at least once per hour q
until either:
1 a)
The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)
THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
2.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:
a)
Reduce the QUADRANT POWER TILT RATIO to within its limit, or b)
Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
3.
Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to SE% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%
or greater RATED THERMAL POWER.
b.
With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
1.
Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
"See Special Test Exceptions Specification 3.10.2 DIABLO CANYON - UNITS 1 & 2 3/4 2-18 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) a)
The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)
THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
2.
Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes; 3.
Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%
or greater RATED THERMAL POWER.
c.
With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod, calculate the QUADRANT POWER TILT RATIO at least once per hour until THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER and:
1.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL I
POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 2.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION j
above 50% of RATED THERMAL POWER may proceed provided that the j
QUADRANT POWER TILT RATIO is verified within its limit at least I
once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.
l d.
The provisions of Specification 3.0.4 are not applicable to POWER OPERATION above 50% of RATED THERMAL POWER.
j l
l i
DIABLO CANYON - UNITS 1 & 2 3/4 2-19 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS i
4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
a.
Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.
Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel in-operable by using the movable incore detectors to confirm that th~e normalized symmetric power distribution, obtained from the 4 pairs of symmetric thimble locations, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
i i
l l
DIABLO CANYON - UNITS 1 & 2 3/4 2-20 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS r3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION
'3.2.5 The following DNB related parameters shall be maintained within the
-limits shown on Table 3.2-1:
Reactor Coolant System T,yg, and a.
b.
Pressurizer Pressure.
APPLICABILITY:
MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to' within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
l
. SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
DIABLO CANYON - UNITS 1 & 2 3/4 2-21 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
L 6
TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Actual Reactor Coolant System T
< 584.3 F (Units 1 and 2 Cycle 4 avg and after)
< 582*F (Unit 2 Cycle 3)
Actual Pressurizer Pressure
> 2212 psia * (Units 1 and 2 Cycle 4 and after)
> 2220 psia * (Unit 2 Cycle 3)
Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL l
POWER.
1 1
DIABLO CANYON - UNITS 1 & 2 3/4 2-22 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3 L_-__-_----_---------_--------
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Manual Initiation a.
Safety Injection (ECCS)
N.A.
1)
Feedwater Isolation N.A.
2)
Reactor Trip N.A.
3)
Phase "A" Isolation N.A.
1 4)
Containment Ventilation Isolation N.A.
5)
Auxiliary Feedwater N.A.
6)
Component Cooling Water N.A.
7)
Containment Fan Cooler Units N.A.
8)
Auxiliary Saltwater Pumps N.A.
b.
Phase "B" Isolation 1)
Containment Spray (Coincident with
-SI Signal)
N.A.
2)
Containment Ventilation Isolation N.A.
c.
Phase "A" Isolation 1)
Containment Ventilation Isolation N.A.
d.
Steam Line Isolation N.A.
2.
Containment Pressure-Nigh a.
SaVety Injection (ECCS) 1 27(1) 1)
Reactor Trip
<2 7 63(4)/28(5) 2)
2)
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3)
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Auxiliary Feedwater
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Component Cooling Water 538
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Containment Fan Cooler Units
< 40 8)
Auxiliary Saltwater Pumps 548(4)/58(5) 3.
Pressurizer Pressure-Low a.
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< 60 1
6)
Component Cooling Water 548
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Containment Fan Cooler Units
< 40 8)
Auxiliary Saltwater Pumps 558(1)/48(4)
DIABLO CANYON - UNITS 1 & 2 3/4 3-28
(
SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS l
LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifica-tions 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
a.
The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b.
The limits of Specifications 3.2.2.1 or 3.2.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specifica-tion 4.10.2.2 below.
APPLICABILITY:
MODE 1.
ACTION:
With any of the limits of Specifications 3.2.2.1 or 3.2.2.2 or 3.2.3 being l
exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 are suspended, either:
a.
Reduce THERMAL POWER sufficient to satisfy the ACTION require-ments of Specifications 3.2.2.1 or 3.2.2.2 and 3.2.3, or l
b.
Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85%
of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during PHYSICS TESTS:
a.
Specifications 4.2.2.1.2 or 4.2.2.2.2 and 4.2.2.1.3 or 4.2.2.2.3, and b.
Specification 4.2.3.2.
DIABLO CANYON - UNITS 1 & 2 3/4 10-2 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
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3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).
APPLICABILITY:
MODE 2.
ACTION:
a.
With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm baron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
b.
With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of-Specification 3.1.1.1.
DIABLO CANYON - UNITS 1 & 2 3/4 10-1
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THIS PAGE INTENTIONALLY LEFT BLANK DIABLO CANYON - UNITS 1 & 2 8 3/4 2-3 Amendment Nos.12and 10
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POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
Each of these is measurable but will normally only be determined periodically as specified in 5,. m Mcations 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to ensure.,nat the limits are maintained provided:
1.
Control rods.in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position s 2.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, 3.
The control rea insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained, and 4.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F will be maintained within its limits provided Conditions 1. through H
4., above, are maintained.
TherelaxationofFfHas a function of THERMAL POWER l
allows changes in the radial power shape for all permissible rod insertion limits.
R, as calculated per Specification 3.2.3 and used in Figure 3.2-3a, Figure 3.2-3b and Figure 3.2-3c accounts for F less than or equal to 1.56 for H
LOPAR fuel and 1.59 for VANTAGE 5 fuel for Units 1 and 2 Cycle 4 and after, andequaltoorlessthan1.49forUnit2Cycge3.
These values are the values used in the various accident analyses where F influences parameters other AH than DNBR, e.g., peak clad temperature, and thus are the maximum "as measured" values allowed.
Margin between the safety analysis limit DNBRs (1.44 ano 1.48 for the LOPAR fuel thimble and typical cells, respectively, and 1.68 and 1.71 for the VANTAGE 5 thimble and typical cells) and the design limit DNBRs (1.33 and 1.37 for the LOPAR fuel thimble and typical cells and 1.50 and 1.32 for the VANTAGE 5 fuel thimble and typical cells, respectively) is maintained for Units 1 and 2 Cycle 4 and after.
A fraction of this margin is utilized to accommodate the transition core DNBR penalty of maximum 12.5 percent and the appropriate fuel rod bow DNBR penalty (less than 1.5 percent for both fuel types per WCAP-8691, Revision 1) for Units 1 and 2 Cycle 4 and after.
The rest of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.
Fuel rod bowing reduces the value of DNB ratio.
Credit is available to offset this reduction in the generic margin for Unit 2 Cycle 3.
The generic margin totaling 9.1% DNBR is derived from the difference between the design and required values on the following items:
(a) design DNBR limit, (b) grid spacing multiplier, (c) thermal diffusion coefficient, (d) DNBR spacer factor mu tiplier and (e) pitch reduction.
The rod bow penalty is calculated with the method described in WCAP-8691, Revision 1, and is completely compensated by the available margin of 9.1%.
DIABLO CANYON - UNITS 1 & 2 B 3/4 2-4 Amendment Nos.37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS l
BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
E When an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made.
An allowance of 5% is appropriate l
for a full core map taken with the incore detector flux mapping system and a l
3% allowance is appropriate for manufacturing tolerance.
When RCS flow rate and F are measured, no additional allowances are l
H necessary prior to comparison with the limits of Figures 3.2-3a, b and c.
Measurement errors of 2.4% for Units 1 and 2 Cycle 4 and after, and 3.5% for Unit 2 Cycle 3forRCStotalflowrateand4%forFfH have been allowed for in determination of the design DNBR values.
For Units 1 and 2 Cycle 4 and after, the 4% is applied to the DNBR limit.
For Unit 2 Cycle 3, the 4% is applied to N
FAH' The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which cou'd lead to operation outside the acceptable region of operation shown on Figures 3.2-3a, b, and c.
For Unit 2 Cycle 3, the Peaking Factor, Fxy(Z), is measured periodically to provide additional assurance that the Hot Channel Facu r, F (Z), remains 9
within its limit.
The F limit for RATED THERMAL POWER (F P) as provided in xy the Peaking Factor Limit Report per Specification 6.9.1.8 was determined from l
expected power control maneuvers over the full range of burnup conditions in the core.
For Units 1 and 2 Cycle 4 and after, the hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power 5
factor appropriate to RAOC operation, W(z), to provide assurance that the limit c., the hot channel factor F (z) is met.
W(z) accounts for the effects 9
of normal operation transients and was detennined from expected power control maneuvers over the full range of burnup conditions in the core.
The W(z) function for normal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.8.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
Tte limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x y plane power tilts.
The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indic ted power tilt.
a DIABLO CANYON - UNITS 1 & 2 8 3/4 2-5 Amendment Nos.37 and 36 Effective at end of Unit 1 Cycle 3
POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Continued)
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the event such action does not correct the tilt, the margin for uncertainty en F is reinstated by reducing the power by q
3% for each percent of-tilt in excess of 1.
3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and safety analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain the DNBR limits throughout each analyzed transient.
l The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
DIABLO CANYON - UNITS 1 & 2 B 3/4 2-6 Amendment Nos.37and 36 Effective at end of Unit 1 Cycle 3
ADMINISTRATIVE CONTROLS
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SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) atmospheric stability.* This same report shall include an assessment of the 5
radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station durhg the previous calendar year.
This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period.
All assumptions used in making these assessments, i.e., specific activity, exposure time and loca-tion, shall be included in these reports.
The. meteorological conditions con-current with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for deter-mining the gaseous pathway doses.
The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION PROCEDURE (0DCP).
The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF TM -UBLIC from reactor releases and other nearby uranium fuel cycle sources. m tuding doses from primary effluent pathways and direct radiatinn, for tN vvious calendar year to show conformance with 40 CFR Part 190, " Environment-
'oeiation Protection Standards for Nuclear Power Operation." Acceptable m.-
c: cr calculating the dose contribution from liquid and gaseous effluent e..pven in Regulatory Guide 1.109, Rev. 1, October 1977.
The Semiannual Radioactive Effluent Release Report
%il include a list and description of unplanned releases from the site to @f,ESTRICTED AREAS of radioactive materials in gaseous and liquid effluents mace during the reporting period.
The Semiannual Radioactive Effluent Release Reports sb include any changes made during the reporting period to the PCP, the EW. and the ODCP, pursuant to Specifications 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.15.
It shall include a listing of new locations for dose cal-culations and/cr environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.
The Semiannual Radioactive Effluent Release Reports sci l also include i
an explanation as to why the inoperability :: liquid or gaseous the following:
effluent monitoring instrumentation was not corrected within the time specified in Specifications 3.3.3.9 or 3.3.3.10, respectively; ar.d description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specifications 3.11.1.4 or 3.11.2.6, respectively.
In lieu of submission with the Semiannua' Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
DIAELO CANYON - UNITS 1 & 2 6-19 I
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4 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.7 Routine reports of operating statistics and shutdown experience, includ-ing documentation of all challenges and failures to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Manage-ment, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
PEAKING FACTOR LIMIT REPORT
6. 9.1. 4 The F limits for RATED THERMAL POWER (FRTP) shall be provided to
~
xy x
the NRC in accordance with 10 CFR 50.4 for all core planes containing Bank "D" control rods and all unrodded core planes at least 60 days prior to each cycle initial criticality.
In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.
This report is not required for the initial cycle.
The W(z) function for Load Follow operation shall be established for at least each reload core and shall be maintained evailable at the plant.
The limits shall be established and implemented on a time scale consistent with normal procedural changes.
The analytical met. hods used to generate the W(z) function shall be those previously reviewed and approved by the NRC.*
If changes to these methods are deemed necessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.
A report containing the W(z) function, as a function of core height (and burnup if applicable), shall be provided to the NRC in accordance with 10 CFR 50.4 within 30 days of their implementation.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering time interval at each power level; b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; c.
All REPORTABLE EVENTS; CWCAP-8385 " Power Distribution Control and Load Following Procedures."
WCAP-9272-A " Westinghouse Reload Safety Evaluation Methodology," and WCAP-10216-P-A " Relaxation of Constant Axial Offset Control /F Surveillance q
Technical Specification."
DIABLO CANYON - UNITS 1 & 2 6-20 Amendment Nos. 37 and 36 Effective at end of Unit 1 Cycle 3
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