ML17310B331
| ML17310B331 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/26/1994 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17310B332 | List: |
| References | |
| NUDOCS 9406080186 | |
| Download: ML17310B331 (48) | |
Text
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~o eCy UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN 50-528'ALO:
VERDE NUCLEAR GENERATING STATION UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
76 License No.
NPF-41 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric
- Company, Southern California Edison
- Company, Public Service Company o'f New Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority dated January 20,
- 1994, compl,ies 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
-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.
NPF-41 is hereby amended to read as follows:
9406080i86 '940526
- PDR, ADOCK05000528 P
'PDR
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3.
(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revi'sed through Amendment No. 76, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this license.
APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection
- Plan, except where otherwise stated in.specific license conditions.
This 1-icense amendment is effective as of the date of issuance and must
'be fully implemented prior to the startup from Cycle 5 Refueling Outage.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
Nay 26, 1994 Theodore
'R. quay, Director
-Project Directorate IV-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
.I+I'
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
76 TO FACILITY OPERATING LICENSE NO.
,DOCKET NO.
STN 50-528 Replace the following pages of the Appendix A Technical 'Specificati'ons with the enclosed, pages.
The revised pages are identi.fied by amendment, number and contain vertical 1'ines indicating, the areas of change.
Remove 2-1 2-3 B 2-1 B 2-2 B 2-6 6-20a Insert 2-1'-3 8 2-1 B 2-2 B 2-6 6-20a
igt
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 1 SAFETY'IMITS
- 2. 1. 1 REACTOR CORE DNBR 2.1.1.1 The. calculated DNBR of the, reactor core shal.l, be, maintained greater than or equal to 1.30.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the calculated DNBR of,the reactor has decreased to less than 1.30, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specifi-cation 6.7. 1.
PEAK LINEAR HEAT RATE 2; 1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21 kW/ft.
APPLICABILITY:
MODES 1 and 2.,
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. 1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System;pressure shall not.exceed 2750 psia.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia,.be in HOT
,STANDBY with: the Reactor Coolant System.pressure within its limit within 1
- hour, and comply.with the requirements of Specification 6.7.. 1.
MODES 3, 4, and 5:
Whenever
.~he Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant'ystem pressure to within its limit within 5 minutes, and comply with the requirements of Spec'ification.6.7.1.
PALO VERDE UNIT 1
2-'1 AMENDMENT NO. ~,
76
SAFETY LIMITS AND LIHITINIG SAFETY SYSTEH SETTINGS
- 2. 2 LIHITING SAFETY S'STEM SETTINGS REACTOR TRIP SIETPOINTS II 2.2. 1 The reaIctor protective instrumentation setpoints shall be set consistent with the'Trip,Setpolnt values shown in Table 2.2-1.
APPLICABILITY:
As shown for each channe'I in,Table 3.3-1.
ACTION:
Vith a reactor protective instrumentation setpoint less Iconservative than the value shown in the Allowable Values column of Talble 2.,2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.
1 until the channel. is res'tored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint,value.
PALO VEROE - IJNIT 1
TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS
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(
m C7m m
C)
FUNCTIONAL UNIT 1.
TRIP GENERATION A.
Process 1.
Pressurizer Pressure - High 2.
Pressurizer Pressure
- Low 3.
Steam Generator Level Low 4.
Steam Generator Level - High 5.
Steam Generator Pressure - Low 6.
Containment Pressure
- High 7.
Reactor Coolant Flow - Low a.
Rate b.
Floor c;
Band I
8.
Local Power Density - High 9.
DNBR - Low B;
Excore Neutron Flux Variable OverpoHOr Trip a.
Rate b.
Ceil~ing c.
Band TRIP SETPOINT s 2383 psia
> 1837 psia (2)
~ 44.2/o (4)
~ 91.0% (9) 2 919 psia (3) s 3.0 psig s 0.115 psi/sec (6) (7)
> 11.9 psid (6)(7) s 10.0 psid (6)(7) s 21.0 kW/ft (5)
~ 1.30 (5) 410.6%/min of RATED THERMAL POWER (8)
<110.0% of RATED THERMAL POWER (8)
~9.7% of RATED THERMAL POWER (8)
ALLOWABLE VALUES
< 2388 psia
> 1821 psia (2)
~ 43.7% (4)
~ 91.5% (9)
> 911 psia (3) s 3.2 psig e
< 0.118 psi/sec (6)(7)
> 11.7 psid(6)(7) 5 10.2 psid (6)(7)
< 21.0 kW/ft (5)
> 1.30 (5)
<11.0%/min of RATED THERMAL POWER (8)
~111.0% of RATED THERMAL POWER (8)
-9.9/. of RATED THERMAL POWER (8)
TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE IHSTRINEHTATIOH TRIP SETPOIHT LIMITS FUNCTIOHAL UHIT
,2.
Logarithmic Poser level -'igh (I) a.
Startup and Operating h
Chait dram i.
Core Protection ialculator System 1.
CEA'Calculators 2.
Core Protection ialculators D.'uppielentary Protection System Pressurizer-Pressure - High RPS LOGIC A:
matrix Logic B
Initiition Lonic III. RPS ACTUATIOH DEVICES A.
Reactor Trip Breakers B.
Hanuil'Trip TRIP SETPOIHT
< 0.01lC of RATED THERHAL POWER
< 0 01'f RATED THERHAL POWER Hot Applicable Hot Appli cable
< 2h69 psia Hot Applicable Hot Applicable Hot Applicable Hot Applicable ALLOWABLE VALUES
< 0.Olla of RATED THERHAL POWER
< 0.011K of RATED THERHAL POWER Hot Applicable Hot Applicable
< 2hlh psia Hot Applicable Hnt Applicable Hot App'l,icable Hot Applicable
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2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which woul'd result in the release of fission products to the reactor coolant.
Overheating of the.fuel cladding is prevented by (1) restrict'ing 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, and (2) main-taining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kW/ft which will not cause fuel centerline melting in any fuel rod.
- First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the actual heat.flux at that location, is indicative of the margin to DNB.
The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.30 based upon a statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.
The DNBR limit of 1.30 includes a rod bow compensation of 1.75M on DNBR.
- Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center),
fuel rod integrity would be maintained only if the design and operating conditi'ons are appropriate throughout the l.ife of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accom-modation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics (lags),
the directly indicated linear heat rate is dynamically adjusted by the CPC program.
PALO VERDE UNIT 1
B 2-1 AMENDMENT NO. +4 76
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I SAFETY LIMITS AND LIMITING SAFETY SY'TEM SETTINGS BASES Limiting Safety System Settings for the Low DNBR, High Local Power
- Density, High Logarithmic Power Level, Low Pres;suIrizer Pressure and High 'Lihea'r Power Level trips, and Limiting Conditions for Operation on,DNBR and kM/ft margin are specified such that there is a high'degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipatled operatidnall occurrences.
2.1.2 REACTOR COOLANT-SYSTEM PRESSURE The restriction of this Safety ILimit protects t'e integrity of the Reacto'r Coolant System from overpressurizatiion and thereby prevents the release o'f
'adionuclides contained in tire reactor coolant from 'reaching the containment atmosphere.
The Reactor Coollant System components are
~desI ig~ned to Section III, 1974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear.
Power Plant Components which pern>its a, maximum transient'ressure of 110% (2750 psia) of'esign pressure.
The Safety Limit of 2750 psiai is therefore consistent With the design criteria and associated code requ~irements.
The entire Reactor Coolant System is hydrotested at 3125-psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Sietpoints specified in'a'blI 2'.2-1
'are the yatues at which the Reactor Trips are
. et 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 tc> assist the Engineered
'afety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set le. s conservative tha'n its'rip Setpoint but wi'thin its specified Allowable Value is acceptabl,e dn 'th5 b'asis that the difference betwe'en each Trip Setpoint and the, A'llowable Value is equal to or less than the
-drift allowance assumed for each trip in the safety analyses'he DNBR - Low and Locall Power Density 1-High a'e-'digit'al'lly generated iirilp setpoints based on Safety Limits of 1.30 and 21 kW/ft, respectiively.
Since these trips are digitally generated by the Core Protection Calculatorsa t'e'rip values are not subject to drifts common to trips generated by ana'log'ype' equipment.
The:Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR, - Low and Local Power Den'sity
-'igh trips include the measurement,,
calculational and processor uncertainties and dynamic allowances as defined in the latest applicable revision of CEN-305-P, "Functional Design Requirements for a Core Protection Calculator,"
and CEN-304-P, "Functional Design Requirements for a Control Element Assembly Calculator.'-'ALO VERDE UNIT 1 AMENDMENT NO. 44 76
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BASES Local Power Densit
- Hi h (Continued) a.
Nuclear flux power, and axial power distribution from the excore flux monitoring system; b.
Radial peaking factors from the position measurement for. the CEAs;.
c.
Delta T power from reactor coolant temperatures and coolant flow
.measurements.
The local power density (LPD), the trip variable, 'calculated by the CPC incorporates uncertainties and dynamic compensation routines.
These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPO after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit.
CPC uncertainties related to peak LPO are the same types used for DNBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density),
sensor time delays, and protection system equipment time delays.
DNBR -
Low The ONBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences.
The DNBR - Low trip incorporates a low pressurizer pressure floor of 1860 psia.
At this pressure a
DNBR - Low trip will automatically occur.
The DNBR is calculated in the CPC utilizing the following information:
a.
Nuclear flux power and axial power distribution from the excore neutr on flux monitoring system; b.
Reactor Coolant System pressure from pressurizer pressure measurement; c.
Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; d.
Radial peaking factors from the position measurement for the CEAs; e.
Reactor coolant mass flow rate from,reactor coolant pump speed; f.
Core inlet temperature from reactor coolant cold leg temperature measurements.
PALO VERDE - UNIT 1 B 2-5 jgggHENT.NL 24
SAFETY'IMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES DNBR Low (Continued)
The DNBR, the trip variable, ca'Iculated by tlhe CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is initiated prior to violation of 'fuel design limits.
These uncertainties and dynamic:
compen. a-tion routines ensure that a reactor trip occurs when the calculated core DNBR
'is sufficiently greater than 1.30 such that the 'decrease in calcu'Iated core DNBR-after the trip wil.l not result'in a vio'lation of the DNBR Safety Limit.
CPC uncertainties related to DNBR cover CPC input'measurement uncertainties, algorithm modelling uncertainties, and. computer equipment processing uncer-tainties.
Dynamic compensation is provided in the CPC. calculations for the effects of coolant tran. port de'I'ays, core heat flux delays (relative to c'hanges in core power), sensor time delays, and protection system equipment time delays.
The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a
CPC initiated trip.
a.
b.
c ~
d..
e.f.
9 Parameter RCS Col.d Leg Temperature-Low RCS Cold Leg Temperature-High Axial Shape Index-Positi've Axial Shape Index-Negati've Pressurizer Pressure-Low Pressuri zer Pressure-Ili glh Integrated Radial Peaking Factor-Low Integrated IRadial Peaking Factor,-High guality Marigin-Low Likikinh Value 70I F
< CId'F
'Not more positive than
+
Not more negative than-
> 1860 psia
< 2388 psia
> 1-.28
< 7.00
>0 0.5'.5'team-Generator Level Hiqh The Steam Genera'tor Level -.High trip is provided to protect the turbine from excessive moisture carry over,.
Since the turbine is automatically tr'ipped when the reactor is, tripped, this trip provides a reliable means for providing protection to the turbine f'rom excesssive moisture carryover.
-This trip's setpoint does not correspond
'to a safety limit, and provides protection in tlhe event of excess feedwater f'low.
The setpoint'~ identical to the main steam isolation setpnint.
Its functiona'll capabi'lity at-the specified trip setting enhances the overall reliabil'ity of the reactor protection system.
PAlO VERDE UNIT I B 2-6 AMENDMENT NO. A4 76
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6'. 1.9 Core operating limits shall be established:
and documented. in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
a.
Shutdown Margin K., - Any CEA Withdrawn for Specification 3. 1 1.2 b.
Moderator Temperature Coefficient BOL and EOL limits for Specification
- 3. 1. 1.3 c.
Boron Dilution Alarms for Specification 3.1.2.7 d.
Movable Control Assemblies - CEA Position for Specification
- 3. 1.3.
1'.
Regulating CEA Insertion Limits for Specification 3. 1.3.6 f.
Part Length CEA Insertion Limits for Specification 3.1.3.7 g.
Linear Heat Rate for Specification 3.2. 1 h.
Azimuthal Power Tilt - T, for Specification 3'.3 i.
DNBR Margin for Specification 3.2.4 j.
Axial Shape Index for Specification 3.2.7 6.9. 1. 10 The.analytical methods used to determine the core operating limits shall. be those previously reviewed and approved by the NRC in:
a
~
b.
C.
d.
"CE Method for Control Element Assembly Ejection Analysis,"CENPD-0190-A, January 1976 (Methodology for Specification 3. 1.3.6, Regulating CEA Insertion Limits).
"The ROCS and DIT Computer Codes for Nuclear Design,"
CENPD-266-P-A, April 1983 (Methodology for Specifications
- 3. 1. 1.2, Shutdown Margin K., Any CEA Withdrawn; 3. 1. 1.3, Moderator Temperature Coefficient BOL and EOL limits and 3..1.3.6, Regulating CEA Insertion Limits).
"Safety Evaluation 'Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No.
STN 50-470, "NUREG-0852 (Novenber 1981),
Supplements No.
1 (March 1983),
No.
2 (September 1983),
No.
3 (December 1987)
(Methodology for Specifications
- 3. 1. 1.2, Shutdown Margin Kg,]
Any CEA Withdrawn;
- 3. 1. 1.3, Moderator Temperature Coefficient BOL and EOL limits;
- 3. 1.2.7, Boron Dilution Alarms; 3. 1.3. 1, M'ovable Control Assemblies-CEA Position;
- 3. 1.3.6, Regulating CEA Insertion Limits; 3. 1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Tilt - T,).
"Modified Statistical Combination of Uncertainties,"
CEN-356(V)-P-A Revision 01-P-A, Hay 1988 and "System 80 Inlet Flow Distribution,"
Supplement 1-P to Enclosure 1-P to LD-82-054, February.1993 (Methodology for Specification 3.2.4, DNBR Margin and 3.2.7 Axial-Shape Index).
PALO VERDE UNIT 1
6-20a AMENDMENT NO. ~, ~
76
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPOR'T (Continued) e.
"Calculatfonal Hethods for the CE Large Break LOCA Evaluation Model,',"
'ENPD-132-PAugust 1974 (Hethddology for Specification 3.2.1, Lfneitr Heat Rate).
f.
"Calculatforial Hethods for the CE Large Break ILOCA Evaluation Model,",.
CENPD-132-P, Suppleii!ent 1, February 1975 (Methodology ',for Specfficatic!n 3.2.1,,
L;inear Heat Rate) g.
"Calculatfonal Methods for the CE Large Break LOCA Eva'luatfon Model)"
'ENPD-132-P, Svpple!r!ent 2-P, July 1975 (Hethodology fo>r Speciffcatfc!n 3.2.1, Lineair Heat Rate).
h.
"Calculative Methods for 'the CE Small Break LOCA Evailuatfon Hodel',"
'ENPD-137-P, August 1974 (Methodo1ogy for Specification 3.2.1, Lfnedr Heat Rate,).
"Calculative Methods.for the CE Smal'1 Break LOCA IEvaluatfon Hodel',"
'ENPD-137-P, Supp'lement 1P, January I1977 (Melthodology for Specfffdatfod 3.2.1, Linear Heat Rate).
j.
Letter:
O'.
D., Parr (NR'C) to F.
M. Stern (CE), dated June 13, 1975 (NRC Staff Review of. the Comlbustfon Engfnhe!'fng ECCS Erealuatfon Model).',
NRC approval for:
6.9.1;10e, 6.,'9.1.10f,'.'9.1.1Oh.'.
Letter:
O.
ID. Parr (NRC). to.A.,E. Scherer (CE), dated December 9 1975 (NRC Staff Review of =the Proposed Combustion Engfneerfng ECCS Evaluatf6n Model Changes).
NRC approva"I for: 6'.9.1.10g.
1.
Letter:
K. Kniel (NRC) to A,. E. Scherer (CE), dat,ed September 27,,
1977 (Eva1luatfon of Topfcail Reports CENPD.133, Supplement 3-P and CENPD-137,,
Supplement 1-P).
NRC approval for 6.9.1.10.f.
The core operating l,imits shall be.determined 0o IthaIt hll applicable limits (e.g., fuel thermal-mechanical limits, core therm'al"'hydraulic limits, ECCS lfm'its, nuclear limits such as shutdown margin;,
and trans'ferlt and'nalysis limits) of the safety analysis are met.
The CORE OPERATING L,IMITS REPORT; including any mid-cycle revisions or supplements thereto, shall be provfd'ed upon is! vance, f'r each re'load cycle, to the NRC Document Control Desk with c'opfes to tlhe Regional Admfnfstrator a'nd'esident Inspector.
PALD VERDE - UNIT 1 6-20b AMENDMENT NO.
69
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN '50-529
, PALO VERDE NUCL'EAR GENERATING STATION UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 62 License No.
NPF-51 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric
- Company, Southern California Edison
- Company, Public Service Company of New Mexico, Los Angeles Department of Water and
- Power, and Southern California Public Power Authority dated January 20,
- 1994, 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 Part I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 'Commission's regulations; D.
The issuance of this, amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment.=:.o this license amendment, and paragraph 2.C(2) of Facility Operating License No.
NPF-51 is hereby amended to read as f llows:
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3.
(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in.Appendix A, as revised through
'Amendment No.62
, and'he Environmental Protecti'on Plan contained in Appendix B, are hereby incorporated into this license.
APS. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection
- Plan, except where otherwise stated in specific license conditions.
This license amendment is effective as of the date of i.ssuance and must be fully implemented prior to the startup from Cycle 5 Refueling Outage.
FOR THE NUCLEAR REGULATORY COMMISSION I
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 26, 1994 Theodore R. quay, Director Project Directorate IV-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
II
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
62 TO FACILITY OPERATING LICENSE NO.
NPF 51 DOCKET NO.
STN 50-529 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove 2-1 2-3 B 2-1 B 2-2' 2-6
.6-20 a Insert 2-1 2-3 8 2-1 B 2-2 B 2-6 6-20a
41
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 1 SAFETY 'LIMITS 2.1.1 REACTOR CORE DNBR
- 2. 1. 1. 1 The calculated DNBR of the reactor core shall be mai'ntained greater than or equal to 1.30.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the calculated DNBR of the reactor has decreased to less than 1.30, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specifi-cation 6'. 1.
PEAK LINEAR HEAT RATE
- 2. 1. 1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21 kW/ft.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the peak linear, heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded'1 kW/ft, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. 1.
REACTOR COOLANT SYSTEM PRESSURE
- 2. 1.2 The Reactor Coolant System pressure shall not exceed'750 psia.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1
- hour, and comply with the requirements of Specification 6.7. 1.
MODES 3, 4, and 5:
Whenever the Reactor-Coolant System pressure has exceeadd'750
- psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements'mf'Specification 6.7. 1.
PALO VERDE UNIT 2 2-1 AMENDMENT NO. ~,
62
SAFETY LIMITS AND LIMITING SAFETY SY'TEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2. 1 The reactor protective instrumentation isetpoirits shall be set consiste'nt'ith the Trip Setpoint va'ilues shown in Table 2'2-1.'PPLICABILITY:
As, shown for each channel in Table 3.3-1.
ACTION:
With a reacto~ protective instrumentation setpoint les's c'onservative than the value shown in the Allowable Values column of Table 2.2-1, declare the c'harine'l inoperable and apply the applicable ACTION 4talteaient r'equirement of Specification
- 3. 3. 1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip 'Setpoint valUe.,'ALO VERDE - UNIT 2 2 2
TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS
(
m C)m m
C7 m
FUNCTIONAL UNIT 1.
TRIP GENERATION A.
Process 1.
Pressurizer Pressure High 2.
Pressurizer Pressure
-. Low 3.
Steam Generator Level Low 4.
Steam Generator Level High 5.
Steam Generator Pressure - Low 6.
. Containment Pressure High 7.
Reactor Coolant. Flow Low a.
Rate b.
Floor c.
Band 8.
Local Power Density High 9.
DNBR Low B.
Excore Neutron Flux l.
Variable Overpower Trip a.
Rate b.
Ceiling c,
Band TRIP SETPOINT s 2383 psia
~ 1837 psia (2)
~ 44.2% (4) s 91.0% (9)
> 919 psia (3) s 3.0 psig s 0.115 psi/sec (6)(7)
> 11.9 psid (6)(7) s 10.0 psid (6)(7) s 21.0 kW/ft (5)
~ 1.30 (5) s10,6%/min of RATED THERMAL POWER (8) sll0.0% of RATED THERMAL POWER (8) s9.7% of RATED THERMAL POWER (8)
ALLOWABLE VALUES s 2388 psia a
1821 psia (2)
> 43.7% (4) s 91.5% (9)
> 911 psia (3) s 3.2 psig s 0.118 psi/sec (6)(7)
> 11.7 psid(6)(7) s 10.2 psid (6)(7) s 21. 0 kW/ft (5)
~ 1.30 (5) sll.0%/min of RATED THERMAL POWER (8) sill.0% of RATED THERMAL POWER (8) s9.9% of RATED THERMAL POWER (8)
'e C)
~C m.
t'ai I
C M
TABLE 2.&l (Continued)
REACTOR PROTECTIVE IHSTRUHENTATIOH TRIP SETPOINT LIHITS ALLOMABLE VALUES ai Startup at4 Operating c h hf le of RATES TNERHAL P(NER
< n 011K of RATED TIIERHAL POMER FUNCTIONAL UNIT
.TRIP SETPOINT 2.
Logarftbefc Peer Level - High (1) mK m'.
5hutdoIm C.
Core Protection Calculator Systea 1.
CEA Ca lculators 2.
Core Protection Calculators 0.
Suppleeentary Protection Systole Pressurizer Pressure - Hfgh i
II.
RPS LOGIC A
Matrix Lodfc S.
Initiation Logic Ill RPS ACTUATIOH DEVICES A,
Reactor Trip Sreakers 8.
ffanu)IITrip U.'ULL% OI KAICU TllCDUSI DAUSV
~ ISa.ns an a
~ vn a.aa U I kaa1Ir sLfa l1V4 IlgwuI ~l Ou S L Hot Applfcable
< 2409 psia Hot Applicable HO% ehpp Ilcab le Hot Applicable Hot Applicable UoUllA 0$
KI%ICV TwERKAI DOMER Nnt ann14rabss Not Applicable
< 2414 psfa Not Applicable Hot App.icable
-Hot Applicable Hot Applicable
~ +
2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding 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 (1) 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, and (2) main-taining the dynamically adjusted peak l.ineai heat rate of the fuel at or less than 21 kW/ft which will'ot cause fuel centerl,ine melting in any fuel rod.
g
- First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.
The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.30 based upon a statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.
The DNBR limit of 1.30 includes a rod bow compensation of 1.75X on DNBR.
- Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center),
fuel rod integrity would be maintained'nly if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accom-modation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics (lags),
the directly indicated linear heat rate is dynamically, adjusted by the CPC program.
PALO VERDE - UNIT 2 B 2-.1 AMENDMENT NO. +9-,
62
BASES Limiting Safety System Settings f'r the Low DNBR, High Local Power
- Density, High Logarithimic Power Leviel, Low PrkssjsrizeI P'ressure and'High Linear Power Level trips; and Limiting Coniditions for'peration on.DNBR and kW/ft'argin are specified.
siuch that therie is a high degree of -confidence that the specified', acceptable fuel design limits -are not exceeded during normal operation and design basis anticiipated operational occurrences.
- 2. 1.2 REACTOR COOLANT SYSTi=M PRESSILJRE The restriction of this Safety Limit protects
.the integrity of the Reactor Coolant System from overpiressurization and thereby prevents tlie release of radionuclides contained in the reactor coolant -from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum, ofthe ASME Code for Nuclear Powe'r Plaht
'omponents which permits a maximum transient pressu're 'of 110% (2750 psia) of design pressure.
The Safety Limit iof 2750 psia is therefore consistent with the design, criteria and a,ssociated code requiremknts.
The entire Reactor Coolant System is hydrjotI sted't'1'25 psia to demonstrate integrity prior to initial operationi 2.2. 1
.REACTOR TRIP SETPOINI'S The Reactor Trip,Setpoints specif'ied in Table 2.2-1 are the.values at which the Reactor Trips are set for each funct,iohal uriit.
The Trip Setpoints have been selected to ensure that the reactor, cbre and Reactor Coolant Systemi 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.
Operation with a trip set, less conservative than its Trip Setpoint but within its specified Allowablie Value is acceptable on the basis that the difference between each Trip Setpioint and the Allowable ValIse is iequal'o or less than the drift allowance assumed for each trip in the safety analyses.
The DNBR - Low anted Local Power Density -'High are digitally generated trip setpoints. based on Safety Limits of 1.30 and 21 kW/ft, respectively.
Since these, trips are digita'lly generated by the Coke Protection Calculators, the trip values are not sulbject to drifts common to.trips generated by analog type equipment.
The, Allowalble 'Values for.the e trips are therefore the same as the Trip Setpoint;s; To maintain the margins of safety a.sumed in the safety analyses, the calculations of the trip variables, for the DNBR ~
Lbw 'and Local Power Density High trips include i;he measurement, calculational'and processor uncertaintiesl and dynamic allowanf.;es as defined II> the latest applicable revision of CEN-305-P, "Functional Design Requirements for a Core Protection Calculator,"
and CEN-304-P, "Functional f)esign Requirements for a Control Element Assembly Calculator."
PALO VERDE UNIT 2 B 2-2 AMENDMENT NO. 49., -62
BASES Local'ower Oensit
- Hi h (Continued) a1 b.
C.
Nuclear flux power and axial power distr'."ution from the excore flux monitoring system;.-
Radial peaking factors from the position measurement for the CEAs; Delta T power from reactor coolant temperatures and coolant flow measurements.
The local power density (LPO), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.
These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak.LPD's sufficiently less than the fuel design limit such that the increase in actual core peak LPO after the trip will.not result in a violation of the Peak Linear Heat Rate Safety Limit.
CPC uncertainties related to peak LPD are the same types used for ONBR calculation.
Dynamic compensation for peak LPD is provided for the effects of core fuel center 1.ine temperature delays (relative to changes in power density),
sensor time delays, and protection system equipment time delays.
DNBR" Low The ONBR -
Low trip is,.provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences.
The DNBR -
Low trip incorporates a low pressurizer pressure floor of 1860 psia.
At this pressure a
DNBR - Low trip will automatically occur.
The DNBR is calculated in the CPC uti lizing the following information:
a.
Nuclear flux power and axial power distribution,from the excore neutron flux monitoring system;
.b.
Reactor Coolant System pressure from pressurizer pressure measurement; c.
Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; d.
e.
Radial, peaking factors from the position measurement for the CEAs; Reactor coolant mass flow rate from reactor coolant pump speed;-
Core inlet temperature from reactor coolant cold leg temperature measurements.
PALO YERDE - UNIT 2 B 2-5 AMENDMENT NO.
SAFETY LIMITS AND LIMITINGSAFETY SYSIEMS SETTINGS.
BASES DNBR - Low (Continued)
The DNBR, the trip variable, calculated by the CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.
These unkerjtaintiies and dynamic compe'nsa--
tion routines ensure that a reactor trip occurs when the calculated c:ore DNBR is sufficiently greatjer tham 1.3i0 such that the decrease in calculated core DNBR after the trip will not result in a violation of the DNBR'Safety Limit.
CPC uncertainties related
'to DNBR cover CPC input m'easurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncer-tainties.
Dynamic compensation is provided in the CPC calculations f'r the effects of coolant transport delays, core heat flux delays (relative to changes in core power),
sensojr time delays, and protection system equipment time delays.
The DNBR algorithm usef. in the CPC is vatic otjIlyi within the limits indicated below and operati'an outside of these limIits will result in: a-CPC initiated trip.
Pajr ametc!r a.
RCS Cold Leg Temperature-Low b.
RCS Cold l.eg Temperature-High c.
Axial Shape Index-Posi tive d.
Axial Shape Index-Negative e.
Pressurizer Pressure-Liow f.
Pressurizer Pressure-High, g.
Integrated Radial Peaking Factor-Low.
h.
I'ntegrated Radial P<!aking Factor-High i.
gual ity MarIyin-Lena
~iiti~nValue
> 470"F
< 610"F Not more positive than
+ 0,.5 Not more negative than 0.5
> 1860 psia
< 2388.psia,
,> 1.28
< 7.00 0'team Generator Level -
~Hi jh The Steam Generator Level; High trip 'is provideiJ to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excesssive moisture carryover.
This; trip's setpoint does not correspond to' safety limit, and provides protection in the event of exc;ess feedbtater flow..
'The setpoint Is identical to the main steam isolation setpoint.
Its fjunctional capability at the specified trip setting enhances the overall reliability of the reactor protection system.
PALO VERDE - UNIT 2',
B 2-6 AMENDMENT NO. t9-62
~ r ADMINISTRATIVE CONTROLS
,I CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
a.
b.
-C.
d.
e.f.
g.
h.l.j ~
Shutdown Margin K., - Any CEA Withdrawn for Specification 3. 1. 1.2 Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.1.3 Boron Dilution Alarms for Specification 3. 1.2.7 Hovabl'e Control Assemblies -
CEA Position for Specification 3. 1.3. 1 Regulating CEA Insertion Limits for Specification 3. 1.3.6 Part Length CEA Insertion Limits for Specific'ation 3. 1.3.7 Linear Heat Rate for Specification 3.2. 1 Azimuthal Power Tilt - T, for Specification 3.2.3 DNBR Margin for Specification 3.2.4 Axial Shape Index for Specification 3.2.7 6.9.1. 10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
a
~
b.
C.
d.
"CE Method for Control Element Assembly Ejection Analysis,"
CENPD-0190-A, January 1976 (Methodology for Specification 3. 1.3.6, Regulating CEA Insertion Limits).
"The ROCS and DIT Computer Codes for Nuclear Design,"
CENPD-266-P-A, April 1983 (Hethodology for Specifications
- 3. 1. 1.2, Shutdown Margin K., - Any CEA Withdrawn; 3. 1. 1.3, Moderator Temperature Coefficient BOL and EOL limits and 3. 1.3.6, Regulating CEA Insertion Limits).
"Safety Evaluation Report related to the Final Design of the Standard.
Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No.
STN 50-470,,
"NUREG-0852 (Novenber 1981),
Supplements No.
1 (March 1983),
No.
2 (September 1983),
No.
3 (December 1987)
(Methodology for Specifications
- 3. 1. 1.2, Shutdown Margin K., 'Any CEA Withdrawn;
- 3. 1. 1.3, Moderator Temperature Coefficient BOL and EOL limits;
- 3. 1.2.7, Boron Dilution Alarms; 3. 1.3. 1, Movable Control Assemblies-CEA Position;
- 3. 1.3.6, Regulating CEA Insertion Limits; 3. 1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Tilt - T,).
"Modified Statistical Combination of Uncertainties,"
CEN-356(V)-P-A Revision 01-P-',
Hay 1988 and "System 80 Inlet Flow Distribution,"
Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2.4, DNBR Margin and 3.2.7 Axial Shape Index).
PALO VERDE UNIT 2 6-20a AMENDMENT NO. M, 62
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS IREPOR1 (Continued) e.
"Calculational Methods for the CE Large Break'LOCA Evaiuazion Hoaj.i,<
CENPD-132-.P, August.1974 (Methodology for Specification-3.,2.1,
',Lihea'r Heat Rate).,
f.
"Calculational Methods for the CE Large Breajk LOCA Evaluation Hodkl,~'ENPD-132-P, Supplement 1, February 1975 (Hethodology for Specification 3.2.1, Lineair Heat Rate).
. g.
"Calculational IMethods for the CE Large Breajc LOCA Evaluation Hoddl,>'ENPD-1'32-P; Supplement 2-P, July 1975 (Methodology f'r Specification 3.2.1, ILinear Heat, Rate).
h.
"Calculative Methods for the CE Sttjiall Break LOCA Evaluation Hodel,,"
CENPD-137-P, August 1974 (Methodology fior Specification 3.2.1, Linear Heat Rate).
"Calculative Methods for the CE Sm'all Bjreak LOCA Eval. ation Hodje1,"
CENPD-137-P, Supplement 1P, January 197'7 (Methodology for Specification 3.2. 1, Linear Heat Rate).
j.
Letter:
0.
D. Parr (NRC) to F.
M. Stern (CE), dated June 13 1975 (NRC Staff Review of the Combustion Eng~ineering'CCS Evaluation Model).'RC'pprova11 for:
6.9.1.10e, 6.9.1.,10f, 6.9.1.1Oh.
k.
Letter:
- 0. D.,Parr (NRC) to A; E. Scherer (C',E),, dated December'9, 1975'NRC Staff Review of the Proposed ICodbuktion Eng~ineering ECCS Ekaluattioh Model Changes).
NRC approval fdr: 6.9.3..1Og.
l.
Letter:.
-K. Kniel (NRC) to A.
E. Scherer (CE), dated September 27,'977 (Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).
NRC ajtpr'oval for 6.9.1.10.i.
The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCi limits, nuclear limits such as shutdown margin, and transient and analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid"cycle revisions or supplements
- thereto, shall be provided upon iss'vance',
for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident. Inspector.
PALO VERDE - UNIT 2 2Ob AMENDHI.:NT NO. 55
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205S~001 ARIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO.
STN 50-530 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.
3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
48 License No.
NPF-74 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric
- Company, Southern California Edison
- Company, Public Service Company of New Mexico, Los Angeles Department of Water and
- Power, and Southern California Publi'c Power Authority dated January 20,
- 1994, complies. with the standards and'equirements of the Atomic Energy Act of 1954, as amended'the Act) and the Commission'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 rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common
.defense and security or to the health and safety of the public;. and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been. satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi'cations as indicated in the.atta'chment to this,license amendment, anJ paragraph 2.C(2) of Facility Operating License No. NPF-74 is hereby
~
amended to read as follows:
igt
3.
(2) Technical S ecifications and E~'vironmenta1 Protection Plan The Technical Specifications contained'n Appendix A, as revised through Amendment No. 49, and 'the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.
APS shall, operate the facility in accordance with the Technical Specifications and. the Environmental Protection
- Plan, except where otherwise stated i'n specific license conditions.
This. license, amendment is effective as of the date of issuance and must be fully implemented prior to the startup from Cycle 4 Refueling Outage.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
gay p6, igg4 Theodore R. quay, Director Project Directorate IV-3 Division of Reactor Projects I'II/IV Office of Nuclear Reactor.
Regulation
4i
~
~
~i ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
48 TO FACILITY OPERATING L'ICENSE NO. NPF-74 DOCKET NO.
STN 50-530 Replace the following pages, of the 'Appendix A Technical Specifications with the enclosed,p'ages.,
The revised pages are identi'fied by amendment number and'ontain vertical lines indicating the areas of change.
Remove 2-1 2-3 8 2-1
.B 2-2 B 2-,6 6-20a Insert 2-1 2-3 B 2-1 B 2-'2 B 2-6 6-20a
~ ~
t
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1
.SAFETY LIMITS l,i.i ~A DNBR 2.'1. l. 1 The calculated DNBR,of the.reactor core shallbe. maintained greater.
than or equal to 1.30.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the calculated DNBR of the, reactor has decreased to less than 1,.30, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specifi-cation 6.7. 1.
PEAK LINEAR HEAT 'RATE l'.
- 1. 1.2 The.peak linear. heat rate (adjusted for fue'1 -rod dynamics) of the fuel shall be maintained less than: or equal'o 21 kW/ft.
APPLICABILITY:
MODES 1 and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT.STANDBY within 1,hour, and comply with the requirements of Specification 6.7. 1.
REACTOR COOLANT SYSTEM, PRESSURE
- 2. 1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICABILITY:. MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever.the.Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within,its limit within 1
- hour, and comply with. the requirements. of Specification"6.7. 1.
MODES 3, 4, and 5:
Whenever the Reactor Coolant 'System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7,. 1.
PALO YERQE. UNIT 3 2-1 AMENDMENT NO. ~,
48
SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SE'TTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2. 1 The reactor protective instrumentatiE)n setpoints shall be set consistent with the Trip Setpoint limits,shown in Tabid 2'.'2-1.'PPI.ICABILITY:
As shown fot each channel in Table 3.3-1.
ACTION:
With a reactor protect'ive instrumentation setpoint less conservative than the
'alue shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specifidation 3.3. 1 until the channel is restored to OPERABLI= status with its trip setp'oint
'djusted consistent with the Trip Setpt)int value.
PALO VERGE - UNIT 3 2"2
TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT 1.
TRIP GENERATION A.
Process 1.
Pressurizer Pressure High 2.
Pressurizer Pressure Low 3.
Steam Generator Level Low 4,
Steam Generator Level - High 5.
Steam Generator Pressure - Low 6.
Containment Pressure
- High 7.
Reactor Coolant Flow Low a.
Rate b.
Floor c.
Band 8.
Local Power Density High 9.
DNBR Low B.
Excore Neutron Flux 1.
Variable Overpower Trip a.
Rate b.
Ceil ing c.
Band TRIP SETPOINT 2383 psia 2 1837 psia (2)
~ 44.2% (4)
< 91,0% (9)
~ 919 psia (3) s 3.0 psig
~ 0.115 psi/sec (6)(7)
> 11.9 psid (6)(7)
< 10.0 psid (6)(7) s 21.0 kW/ft (5)
~ 1.30 (5)
F10.6%/min of RATED THERMAL POWER (8)
~110.0% of RATED THERMAL POWER (8) s9.8% of RATED THERMAL POWER (8)
ALLOWABLE VALUES
< 2388 psia
> 1821 psia (2)
~ 43.7% (4) s 91.5% (9)
> 912 psia (3) s 3.2 psig
< 0.118 psi/sec (6)(7)
> 11.7 psid(6)(7)
< 10.2 psid (6)(7) s 21.0 kW/ft (5)
~ 1.30 (5)
~11.0%/min of RATED THERMAL POWER (8)
<Ill,0% of RATED'HERMAL POWER (8)
<10.0% of RATED THERMAL POWER (8)
TABLE 2. 2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIHITS FUNCTIONAL UNIT 2.
Logarithmic Power Level - High (1) a.
Startup'nd Operating b.
Shutdown TRIP SETPOINT
< 0.01(C of RATEO IHERHAL POWtK
-< 0.010K of RATEO THERMAL POWER ALLOWABLE.VALUES
< 0.011K of RATEO THERHAL POWER
< 0.011K af RATEO THERMAL POWER C.
Core, Protection Calculator Svstea 1.
CEA Calculators 2'.
Core Protection Calculators Not Applicabl e.
Hot Anniicahle Not Applicable A
1~
Ll nva. nppssa.csun@
O.
Suppleaentary Protection System Pressurizer Pressure - Hiah II.
RPS LUGIC A.
Hatrix Logic Initiation Logic III. RPS ACTUATION OEVICES Reactor Trip BreaKers B.
Hanual Trip
<--2409-ysia-Hot Applicable Not Applicable
<--2%4 p>4 Not Applicable Hot Applicable Not Applicable Hot Applicable
2.1. and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES
- 2. 1. 1 REACTOR'ORE
'he restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would'esult in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by (1) 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, and (2) main-taining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kW/ft which will not cause fuel centerline melting in any fuel rod.
- First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed "departure
- from, nucleate boiling" (DNB).
At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the pr edicted DNB heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.
The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.30 based upon a statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.
The DNBR limit of 1..30 includes a rod bow compensation of 1.75%
on DNBR.
- Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level'i.e., with some melting in the center),
fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume changes which accompany the solid to liquid phase change are significant and require accom-modation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics (lags),
the directly indicated linear heat rate is dynamically adjusted by the CPC program.
PALO VERDE UNIT 3 B 2-1 AMENDMENT NO. ~
48
BASES Limiting Safety System. Settings For the Love DNBR', iligh Local Power
- Density, High Logarithmic Power
- Lesiel, Low Pressurizer Pressure and ILigh Linear Power Level trips, and Limiting Conditions foir Opei ation on DNBR and kW/ftj margin are specified such that there is a high clegI ee of confidence that iIhe specified acceptable fuel design l,imits are nbt 'exdee'ded during normal operation and design basis anticipated operatilonlal occurrences.
- 2. 1.2 REACTOR COOLANll S'STEM PRESSURIE The restriction of this Safety L'imit protec'ts thh integr'ity of the Reactor Coolant System from overpressurization arid thiereby prevents the release of
-'radionuclides contained in the reactoir coolant from reaching the containment atmosphere.
The Reactor Coolant Sy'tem components are designed to Sectiion III,
'1974 Edition, Summer 197!5 Addendum, o)F the ASME 'Code for Nuclear Power Plant Components which permits a maximum transient jpress(ire of 110/.
(2750 psia) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initia'll operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip SeIj;points specified in Table 2.2-1 are the 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 Coo'lant System are prevented from exceeding their Safety Limits during normal operation and design basis antici,pated operational occurrences and to assis't the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but witlIiin 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 thati
'the drift allowance assumed fair each trip in the safety analyses.
The DNBR Low and Local Power Density - High are digita'ily generated'rip setpoints based on Safety Limits of 1.30 and 2,1 kW/ft, respectively.
i Since these trips are digitally generated by the Core Protection Calculators, the trip values are -not subject Ij;o drifts common to trips generated by analog type equipment.
The Allowable Values for these trips are therefore the same
- as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR Low and L.ocal Power Density High trips include the measurement, calculatidnati and processor 'uncertainties and dynamic allowances
- a. defined in the latest applicable revision of CEN-305-P, "Functional Design Requirements for a Core Prbteiction'alculator" and CFN-304-P, "Functional l)esign Requirements, for a Contro'1 Element Assembly Calculator."
PALO VERDE UNIT 3 B 2-2 AMENDMENT NO. m; 4S
BASES Local Power Oensit
- Hi h (Continued) a.
Nuclear flux power and axial power distribution from the excore flux monitoring system; b.
Radial peaking factors from the position measurement for the CEAs; c.
Oelta T power from reactor coolant temperatures and'oolant flow measurements.
The local power density (LPO), the trip variable, calculated by the CPC incorporates uncertainties and'ynamic compensation routines.
These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPO is sufficiently, less than the fuel design limit such that the increase in actual core peak LPO after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit.
CPC uncertainties related to peak LPO are the same types used for ONBR calculation.
Oynamic compensation for peak LPO is provided for the effects of core fuel centerline temperature delays (relative to changes in power density),
sensor time delays, and protection system equipment time delays.
ONBR - Low The ONBR -
Low trip is provided to prevent the ONBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences.
The ONBR - Low trip incorporates a low pressurizer, pressure floor of 1860 psia.
At this pressure a
ONBR - Low trip will automatically occur.
The ONBR is calculated in the CPC utilizing the following information:
a.
Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.
C.
d.
e.
Reactor Coolant System pressure from pressurizer pressure measurement; Oifferential temperature (Oelta T) power from reactor coolant temperature and coolant flow measurements; Rad)H peaking factors from the position measurement for the CEAs; Reactor coolant mass flow rate from.reactor coolant pump speed;.-
Core inlet temperature from reactor coolant cold leg temperature measurements.
PALO VEROE - UNIT 3 AMENOMENT NO.
18
SAFETY LIMITS AND LIMITING SAF'ETY SYS'il'EMS SETI INGS, BASES DNBR - Low-(Continued)
The DNBR, the trip variable, calculated Iby the CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is=initiated prior to v'iolation of fue'1 design limits.
lhese uncertainties and dynamic compensa-'tion-routines ensure that a,reactor trip occurs when the calculated core O'NBR
,is sufficiently greater than 1.30 such that the decrease in calculated core f
DNBR after 'the trip will not result in a violati'on 'of'h'e QNBR Safety Limit.
,CPC uncertainties related to DNBR cover CPC input-measurement uncertainties, algorithm modelling uncertaint,ies, and computer 'eqUip5ent processing uncer-tainties.
Dynamic compensation is provided in the 'CPC c'alcul'ations for the effects of coolant transport
- dlelays, core heat flux delays (relative to changes in core power),
sensor time delays, and protection 'syste'm equ'ipment time delays.
.The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limi~ts~will result in a CPC initiated trip.
a
~
b.
C.
d.
e.f.
g.
h.
Paramete',r RCS Cold I eg Temperature-Low RCS Cold leg Temperature-High Axial Shape Index-Positive Axial Shape Index-Negative Pressurizer Pressure-Low Pressurizer Pressure-High Integrated Radial Peaking Factor-l.ow Integrated Radial PeakiIng Factor-High equality Margin-Low Limitina Value
> 470'iF
< 6l10'F Not mdre positive than
+ 0.5 Not mdre negative than 0.5 5 1860 podia' 2388 psia
< 7.00
>0 Steam Generator Level - Hliqh The Steam Generator Level High triIp is provided to protect the turbine from excessive moisture carry over.
Since the turbine is automatically tripped when the reactor is tripped,, this trip provides 4 rel.iable 'means for providing protection to the turbine from excesssive moisture ~carryover.
This trip's~ se~t-
~
point does not correspond to a safety limit, and provides protection in'he event of excess feedwater f'low.
The setpoint is identical to the main steam isolation setpoint.
Its functional capability at the specified trip setting enhances the overal11 reliability of the reactor protection sy. tern.
PALO VERDE - UNIT 3 B 2-6 AMENDMENT NO. W; 48
ADMINISTRATIVE CONTROLS CORE OPERATING. LIMITS REPORT C.
d.
e.f.
g, h.l.j ~
6.9. 1. 10 The analytical methods used to determine the core operating limits shall be those. previously reviewed and. approved by the NRC in:
"CE Method for Control Element 'Assembly Ejection Analysis,"
CENPD-0190-A, January 1976 (Methodology for Specification 3. 1.3.6, Regulating CEA Insertion Limits).
6.9 '.9 Core operating limits shall be established'.and documented in the CORE OPERATING'IMITS REPORT before each reload cycle or any remaining part of a reload cycle for the fol.lowing:
a.
Shutdown Margin K., Any CEA Withdrawn for Specification 3'. 1. 1.2 b.
Moderator 'Temperature Coefficient BOL and EOL limi'ts for
. Specification 3.1.1.3 Boron Dilution Alarms for Speci'fication 3. 1..2.7 Movable Control Assemblies CEA Position for Specification 3. 1.3. 1 Regulating CEA Insertion Limits for Specification 3. 1.3.6 Part Length, CEA Insertion Limits for Specification 3. 1.3.7 Linear 'Heat Rate for Specification 3.2. 1 Azimuthal Power Tilt - T, for Specification 3.2.3 DNBR Margin for Specification 3.2.4 Axial Shape Index For Specification 3.2.7 b.
C.
d.
"The ROCS and DIT Computer Codes for Nuclear Design,"
CENPD-266-P-A, April 1983 (Methodology for Specifications
- 3. 1. 1.2, Shutdown Margin
'K., - Any CEA Withdrawn; 3.1.1.3, Moderator Temperature Coefficient BOL and EOL limits and 3. 1.3.6, Regulating CEA Insertion Limits).
"Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No.
STN 50-470, "NUREG-0852,(Novenber 1981),
Supplements No.
1 (March 1983),
No.
2 (September 1983),
No.
3 (December 1987)
(Methodology for Specifications
- 3. 1. 1.2, Shutdown Margin K., Any CEA Withdrawn; 3..1. 1.3', Moderator Temperature Coefficient BOL and EOL limits;
- 3. 1.2.7, Boron Dilution Alarms; 3. 1.3. 1, Movable Control Assemblies-CEA Position;
- 3. 1.3.6, Regulating CEA Insertion Limits; 3. 1.3.7, Part Length CEA Insertion Limits and 3.2.3 Azimuthal Power Tilt T,).
"Hodi'fied Statistical Combination of Uncertainties,"
CEN-356(V)-P-A Revis.ion Ol-P-A, May 1988 and "System 80'Inlet,Flow Distribution,"
Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2.4, DNBR Margin and 3.2.7 Axial Shape Index).
PALO VERDE - UNIT 3 6-20a AMENDMENT NO.~
48
ADMINISTRATIVE CONTROLS CORE OPERATING LIHITS REPORT, (Contiriued)
,e.
"Calculational Meth!ods for the CL L'arge 'Bre'ak LOCA Evaluation.Model,"'ENPD-..132-P, August 1974; (H!ethodology 'for Specification 3.2.1,, Linear Heat Rate).;
f:
'",Calculational-Methods f(!r. 'the CE Large Break LOCA Ev'alu'ation Model.;"
CENPD-,132-P, Supplement 1', IFebruary 1975 '(Methodolo!jy for.
Specification,3.2.1, Linear Heat Rate).
g.
"Calculational Methods for 'the'E Large;Brelak
'LOCA Fvaluation Model,"
CENPD-.132-P, Supple!hent,2-P, July"1975'..(Methodology for Specification 3.2.1, Li'near Heat I<ate).
h.
"Cal'culative Methods for: the CE.Small. Break LOCA, Evaluation'Model-,"
CENPD-137-.P, August 1974 (Methodology 'for Specification 3.2,.1, Lineai
'Heat Rate).*
"Calculative, Methods for the CE Small'rdak LOCA 'Evaluation Hodel,"
CENPD-137-P, Supplement 1P, January. 1977 '(Meth'odblogy for Sp'ecificatiori 3.2.1, Linear'Heart Rate)..
j.
Letter:
0.
D.
- Parr, (NRC) to F-.,M. ~tern (CF), dated June,13, 1975 '(NRC Staff 'Revie>v of,the 'Combustion Engineelring ECCS Evaluation Model);
NRC'pproval'oi':
6.9. 1.10e, 6.,9. 1.10f', 6';9.1.'.LOh.
k.
Letter:
- 0. 'D. Parr (NRC) to A.
E. Scherer (CE), dated December,9, 1975 (NRC. Staf f 'Re'view of,the, Proposed Cr!mb!ustioh,Engineering
.ECCS Evaluation
'odel Changes).,NRC. alpproval,for;
-Ci; 9. 1. 10g 1.
Letter K'. Kni'el (NRC) to A;, E.,Scherer
'(CL'), dated September
'27~,
'1977 (Evaluation of.-Tolpical.Reports'ENPD-133,,
Supple'ment'3-P,and CENPD-137.,
Suplple'merit 1-.P').
NRC app'r oval for '6.9.1,'10.i.
The core operating limits shall be.-determined so that all app1licable'imits (e;g;,,fuel'hermal-mechanical-1-imits ciore thermal'=hydr'abulic "limits, ECCS delimits,
~
nuclear limits such as shutdown.!iiargir!, and transierit and ana1lysis limits) 'of 'the safety analysis are: met.,
The CORE OPERATING LIMITS REPORT; inc1luding; any-mid-cycle revision's or
!supplements thereto!; -shall be: pi ovided upon* issu'arice, fo'r each reload: cycle,
- to the NRC Document,'.Control Desk with colpies to
- the Regional -Administrator 'a'nd
'Resident Inspector.
PALO VERDE -'NIT 3 6"20b AMENDMENT NO