ML17298B356

Notification of 840927-28 & 1003-05 Meetings W/Util & C-E in Bethesda,Md to Discuss Various Portions of Tech Specs. Marked-up Tech Specs Encl

ML17298B356
Person / Time
Site:	Palo Verde
Issue date:	10/01/1984
From:	Licitra E Office of Nuclear Reactor Regulation
To:	Knighton G Office of Nuclear Reactor Regulation
References
NUDOCS 8410180054
Download: ML17298B356 (48)

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Docket No.:

50-528 QQ)

IQT 1 84'EMORANDUM FOR:

George W. Knighton, Chief Licensing Branch No.

3 Division of Licensing FROM:

SUBJECT:

E. A. Licitra, Project Manager Licensing Branch No.

3 Division of Licensing FORTHCOMING MEETINGS WITH ARIZONA PUBLIC SERVICE COMPANY ON PALO VERDE UNIT 1 TECHNICAL SPECIFICATIONS DATES 5 TIMES:

See enclosed Table.

LOCATIONS:

See enclosed Table.

PURPOSE:

PARTICIPANTS:

To discuss various portion of the Palo Verde Technical Specifications.

(See enclosed portions).

APS Steve Frost, et al.,

CE Representatives as needed.

NRC E. Licitra, J.

Donohew, NRC Branch Representatives

Enclosures:

As stated cc:

See next page E. A. Licitra, Project Manager Licensing Branch No.

3 Division of Licensing hh DL:LBg3 /

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Table Scheduled Meetin e

On Palo Yerde nit ec naca eci ica ions Date and Time

Thursday, September 27, 1984 1:30 pm

Friday, September 28, 1984 9:00 am and 1:30 pm Wednesday, October 3, 1984 9:00 am and 1:30 pm (except 12:00 - 1:30)

Location CE Office Room 1310 Landow Bui 1ding

Bethesda, Maryland NRC P-118 NRC P-110 NRC Branch Involved RSB/DSI CPB/DSI MTEB/DE AEB/DSI CEB/DE METB/DSI RAB/DSI DHFS

Thursday, October 4, 1984 9:00 am and 1:30 pm Friday, October 5,

1984 9:00 am and 1:30 pm NRC MNBB 6507 NRC MNBB 6110 CSB/DSI PSB/DSI SGEB/DE ASB/DSI MEB/DE RSB/DSI

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Comments and Suggestions We have reviewed the above-mentioned sections of the Technical Specifications which pertain to our area of responsibilty and offer the following comments and suggestions:

/1.

Safety Limit 2.1.1.1 Comment:

See item 5.

g 2.

Footnote (5) of Table 2.2'-1 on page 2-5 shoo'Id add the'following sentences:

The approved SCU-equivalent DNBR limit is 1.231 which includes a

rod bow canpensation of 0.8 percent DNBR; If fuel burnup exceeds 20,000 HWD/HTU with a rod bow.penalty greater than 0.85, the DNBR limit should be adjusted.

A DNBR trip setpoint of 1.231 is allowed provided that the difference -is compensated by an increase in the CPC addressable constant BERR1 as follows:

BERR1 = BERR1 ld

[1 + ~RB-O.B x

d POL P

(DNBR)I Where BERR1

'is the uncompensated value of BERRl; RB is the fuel rod old bow penalty in X DNBR; POL is the power operating'imit; d(POL)/d (DNBR) is the absolute value of the most adverse derivative of POL with respect to DNBR.

3.. The CPC type II addressable constants should be listed in Table 2.2-2.

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the end of the'hird paragraph of Bases 2.1.1'n page B 2-1, add the following sentences:

The DNBR Iiml,t of 1.231 includes a rod bow compensation of 0.8 percent on DNBR.

For fuel Mnups exceeding 2D,000 NWD/HTU with a rod bow penalty greater than 0.8 percent DNBR, the'NBR limit should be adjusted.

At the end of the second paragraph of Bases 2.2. 1, add the following sentences:

The DNBR trip setpoint should be increased for fuel rod bow penalty greater than 0.8 percent ONBR.

But the trip setpoint of 1.231 is allowed if the required GNBR. increase is compensated by an increase of the addressable constant BERR1.

Bases 3.2.5 provides

a. value. for t'er miaimum,flowzate required: in the

-,.-'<Reactor,'Coo1ant System but.i.s.missing the val~e for the maximum flaw rate permitted.

7.

There are two items not listed in Table 3.3-10 for II.F.2, Inadequate,...

Core Cooling Instrumentation:

(a)

Reactor Yessel Monitoring System (b)

Core Exit Thermocouples 8.

There are two items not listed in Table 4.3-7 for II.F. 2, 'Inadequate t

Core Cooling Instrumentation:

9.

(a)

Reactor Yessel benitoring System (b)

Core Exi)t Thermocoep'lee In Technical Specification 4.1.1.2.2 it is not clear why a k ff of 0.98 is mentioned whereas the required shutdown margin is w4%b k.

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10.

Ln Technical: Specification 3.1.3.5 th'e fully withdrawn CEA position is referred to as 144.75 inches or'greater.

Has the effect of this CEA bite been included in calculating the physics characteristics such as scram worth's, shutdown margins, peaking factors, etc.?

r ll.

Rhy has thresholdmf above 5X of rated thermal power (STS} been changed to above 20K of rated thermal power ln the surveillance requirements of 3.10. 2 and.3.10.4?

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TABLE 2. 2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1)

Trip may be manually bypassed above 10-~X of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10-4X of RATED THERMAL POWER.

(2)

(3)

Q (a)

(5)

(6)

In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

In MODES 3-6, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

X of the distance between steam generator upper and lower level wide range. instrument nozzles..

As stored within the Core Protection'alculator (CPC).

Calculation of the trip setpoint includes measurement, calculational and processor uncer-

tainties, and dynamic allowances.

Trip may be manually bypassed below 1X of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to Lt'f RATER THERMAL POWER.~

RATE is the maximum rate of decrease of the trip setpoint.

FLOOR is the minimum value of the trip setpoint.

BAND is the amount by which the trip setpoint is below the input signal unless limited by Rate or Floor.

(7)

The setpoint may be altered to disable trip function during testing pursuant to Specification

3. 10.3.

(s)

RATE is the maximum rate of increas'e of the trip setpoint.

There are no restrictions on the rate at which the setpoint can decrease.

CEILING is the maximum value of the trip setpoint.

EEMD >s the amount by which the trip setpoint is above the input signal

~un ess limited by the rate or the ceiling.

X of the distance between steam generator upper and lower level narrow range instrument nozzles.

PALO VERDE - UNIT 1 2"5

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The approved SCU-equivalent DNBR limit is 1-.231 which Anc1udes a

rod bow ccmpensation of.0.8,percent DNBR; If fuel burnup exceeds 20,POO. N'ND/NTU with a rod bow penalty greater than 0.8X, the DNBR limit: should be adjusted.

-A DNBR.trip setpoint of 1.231 is allowed -provided =that the".difference. is..compensated by an increase i'n the CPC addressable constant BERR1'~'as follows:

L BERR1 BERR1 ><

[1'

~RB'-0.8 x

POL 100 (DNBR)

Where BERRl is the uncompensated value of BERR1; RB is the fuel rod old bow penalty in 5 DNBR; POL is the power'perating'imit; d(POL)/d (DNBR) is the absolute value of the most adverse derivative of POL with respect to DNBR.

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TABLE. 2.2-2 Continued I

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS I.

TYPE II ADDRESSABLE CONSTANTS POINT jD PROGRAM NUMBER LABEL DESCRI PTIOQ 68 69 70 71 72

~ 73 75 BERRO BERR1 BERR2 BERR3 BERR4 EOL ARHl ARH2 Thermal power uncertainty bias Power uncer tainty factor used in DNBR calculation Power uncertainty bias used in DNBR calculation Power uncertainty factor used in local power density calculation Power uncertainty bias used in local power density calculation End of life flag Multiplier for planar radial peaking factor

~ Multiplier for planar radial peaking factor 76 ARH3 77 ARM4" Hultiplier-for planar radial peaking

"'Mul'&plierfor'planai radial" peaking factor fa'ctor 78 79 ARH5 ARM6 Multiplier for planar 'radial peaking Multiplier for planar radial peaking factor factor 80 82 83 ARM7 SC11 SC12 SC13 SC21 Multiplier for planar radial peaking factor Shape annealing correction factor Shape annealing correction factor-Shape annealing correction factor Shape annealing coi rection factor 85 86 SC22 Shape SC23 Shape annealing correction factor annealing correction factor 87 SC31 Shape annealing correction factor 88 SC32 Shape annealing correction factor

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TABLE 2.2-2 (Continued)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS I.

TYPE II ADDRESSABLE CONSTANTS Continued POINT ID PROGRAM.

NUMBER

. LABEL DESCRIPTION 89 SC33 Shape annealing correction factor 90 PFMLTD 9l;., PFMLTL DNBR penalty factor correction multiplier

..LPD. penalty..factor correction multiplier 92

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93 ASM2

.Multiplier for CEA shadowing factor ASM3 Multiplier for CEA shadowing factor 95 96 97 98 ASM4 ASM5 ASM6 ASM7 CORRI Multiplier for CEA shadowing factor Multiplier for CEA shadowing factor Multiplier for: CEA shadowing factor Multiplier for CEA shadowing factor Temperature shadowing correction factor multiplier

"'--:-.'99::-"':~--'.,BPPCC1 100 BPPCC2 BPPCC3 BPPCC4 101 102

- " ".-'-:-Boundary'point power correlation c'oundary point power correlation Boundary point power correlation Boundary point power correlation coefficient coefficient coefficient coefficient

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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 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) maintaining 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.

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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 l. 231 based upon a. statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.~~~

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 accommodation.

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.

Limiting Safety System Settings for the Low DNBR, High Local Power Density,

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High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Powe'r

'evel trips, and Limiting Conditions for Operation on DNBR and kW/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 anticipated operational occurrences.

PALO VERDE - UNIT 1 8 2-1

The DNBR limit of 1.231 includes a rod bow compensation of 0.8 percent on DNBR.

For fuel lxlTnups exceeding 20,000 HWD/HTU with a rod bow penalty greater than 0.8 percent DNBR, the DNBR limit should be adJusted.

C SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS

'BASES

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction oQ,this Safety. Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the 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, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110K (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 initial operation.

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2. 2. 1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.

The Trip Setpoints

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have been selected to ensure that the reactor core and Reactor Coolant System

.,:,are prevented from exceeding their Safety Limits during normal operation and

': 'esign 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 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 assumed for each trip in the safety analyses.

The DNBR -

Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of 1.231 and 21 kW/ft, respectively.

Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment.

The Allowable Values or these trips are therefore the same as the Trip Setpoints. ~ +

k, To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density-High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CESSAR System 80 applicable system descriptions and safety analyses.

Manual Reactor Tri

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The Manual reactor trip is a redundant channel to the automatic

'protective instrumentation channels and provides manual reactor trip capability.

PALO VERDE - UNIT 1 B 2-2

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The DNBR trip setpoint should be increased for fuel rod bow penalty.

greater than 0.8 percent ONBR.

But the trip setpoint of 1.231 is allowed i'f the required DNBR increase is compensated by an increase of the addressable

constant, BERR1.

POWER DISTRIBUTION LIMITS m

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LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reactor Coolant SP tern total flow rate shall be greater than or equal to 164. 0 x 10e Ibm/hr.

0 APPLICABILITY:

NODE 1.

(

) lorn/Q ACTION:

With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS

4. 2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than its limigat least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

PALO VERDE - UNIT 1 3/4 2-8

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TABLE 3.3-10 POST ACCIOENT MONITORING INSTRUMENTATION INSTRUMENT REgUIREO NUMBER OF CHANNELS MINIMUM CHANNELS OPERABLE 2.

3.

4.

5.

7.

8.

9.

10.

12.

13.

Containment Pressure Reactor Coolant Outlet Temperature Th t (Wide Range) hot Reactor Coolant Inlet Temperature - T ld (Wide Range) cold Pressurizer Pressure

- Wide Range Pressurizer Water Level Steam Generator Pressure I

Steam Generator Water Level - Wide Range Refueling Water Storage Tank Water Level Auxiliary Feedwater Flow Rate Reactor Cooling System Subcooling Margin Monitor Pressurizer Safety Valve Position Indicator Containment Water Level (Narrow Range)

Containment Water Level (Wide Range) 2/steam generator 2/steam generator 2

1/val ve 1

1/loop 1/loop 1

1/steam generator 1/steam generato'r 1

1 1/val ve Pe.~& Qg~yg Care%

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Q TABLE 4.3-7 POST"ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT 1.

Containment Pressure 2.

3.

4 5.

6.

Steam Generator Pressure 7.

Steam Generator Water Level - Wide Range I

8.

Refueling Water Storage Tank Water Level 9...Auxiliary Feedwater Flow Rate 10.

Reactor Coolant System Subcooling Margin Monitor 11.

Pressurizer Safety Valve Position Indicator 12.

Containment Water Level (Narrow Range) 13.

Containment Water Level (Wide Range) lw. ~+i@i.~ 4wX I%rrvA 5g~~~

Reactor Coolant Outlet Temperature - Th t (Wide Range) hot Reactor Coolant Inlet Temperature

-T ld (Wide Range) cold Pressurizer Pressure - Wide Range Pressurizer Water Level CHANNEL CHECK M

M M

M M

CHANNEL CALIBRATION R

R R

R R

R R

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REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN T ol LESS THAN OR E UAL TO 210 F

LIMITING CONDITION FOR OPERATION

3. 1. 1.2 The SHUTDOWN MARGIN shall be greater than or equal to 4.0X delta k/k.

APPLICABILITY:

MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 4.0X delta k/k, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS

-.. 4. 1. 1.2. 1 The SHUTDOWN MARGIN shall be determined to be greater than or equal M to 4.0X delta k/k:

'a 4 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an'noperable CEA(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.

If the inoperable CEA is immovable or untripp5ble, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or'ntrippable CEA(s).

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

Reactor Coolant System boron concentration, CEA position, Reactor Coolant System average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration,and Samarium concentration.

1.

2.

3.

4.

5.

6.

C.

at least once per 24 fours, when the RCS water level is drained below the pressurizer low level instrument tap, by performing a reactivity balance considering the factors listed in Specification 4. 1. 1.2. lb.

G PALO VERDE - UNIT 1 3/4 1-3

C'PECIAL TEST EXCEPTIONS

... ~l ~lriq! ",i.A'..y 3/4. 10. 2 MODERATOR TEMPERATURE COEFFICIENT GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications

3. 1. 1.3,

3. 1.3. 1, 3. 1.3.5,

3. 1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C. 1 (CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is restricted to the test power plateau which shall not exceed 85K of RATED THERMAL POWER, and b.

The limits of Specification 3.2. 1 are maintained and determined as specified in Specification 4. 10'.2 below.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3; 1. 1.3,

3. 1.3. 1, 3. 1.3.5,

3. 1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C. 1 (CEA Calculators) of Table 3.3-1 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2. 1, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS

4. 10. 2. 1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications

3. 1. 1.3,

3. 1.3. 1,

3. 1.3.5,

3. 1.3.6, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE require-ment of I.C. 1 (CEA Calculators) of Table 3.3-1 are suspended and shall be verified to be within the test power plat~e g, Q %rP(~

4. 10.2.2 The linear heat rate shall e determined to be within the limits of Specification 3.2. 1 by monitoring i continuously with the Incore Detector Monitoring System pursuant to the equirements of Specifications 4.2. 1.3 and 3.3.3.2 during PHYSICS TESTS above of RATED THERMAL POWER in which the requirements of Specifications

3. 1.1.3,

3. 1.3. 1, 3. 1.3.5,

3. 1.3.6, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of I.C. 1 (CEA Calculators) of Table 3.3-1 are suspended.

PALO VERDE - UNIT 1 3/4 10-2

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SPECIAL TEST EXCEPTIONS 3/4.10.4 CEA POSITION REGULATING CEA INSERTION LINITS ANO REACTOR COOLANT C

LD L G EMPERATUR LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.1.3.1, 3.1.3.6 and 3.2.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient, moderator temperature coefficient; and power coefficient provided the limits of Specification 3.2.

1 are maintained and determined as specified in Specification 4.10.4.2 below.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications

3. 1

~ 3. 1, 3. l. 3. 6 and 3.2. 6 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2. 1, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS

4. 10.4.

1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications

3. 1.3. 1, 3. 1.3.6 and/or 3.2.6 are suspended and shall be verified to be within the test power plateau.

4. 10.4.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.

1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specification

3. 3. 3. 2 during PHYSICS TESTS above of RATED THERMAL POWER in which the requirements of Specifications

3. l. 3. 1, 3.

. 3. 6 and/or 3.2. 6 are suspended.

G PALO VERDE - UNIT 1 3/4 10"4

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3/4. 4. 8 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION

3. 4.8.

1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

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a.

A maximum heatup rate of 20'F per hour with the RCS cold leg temperature less than or equal to 95'F, 40'F per hour with RCS cold leg temperature greater than 95'F but less than or equal to 400'F, and 100'F per hour with RCS cold leg temperature greater than 400~F.

b.

A maximum cooldown rate of 20'F per hour with RCS cold leg temperature less than or equal to 100~F, 40'F per hour with RCS cold leg temperature greater than 100'.F but less than or equal. to 130'F, and 100'F per hour with RCS cold leg temperature greater than 130'F.

c.

A maximum temperature change of 10'F in any 1-hour period during inservice hydrostatic and leak testing operations.

APPLICABILITY: At al 1 times.

ACTION:

With any of the. above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T

ld and pressure to less than 210'F and 500 psia, respectively, within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s:

SURVEILLANCE RE UIREMENTS C

4.4.8. l. 1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system

heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.8. 1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR Par~0 Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figure 3.4"2~ ~~/

PALO VERDE " UNIT 1 3/4 4-28

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{1) the actual shift in reference temperature for plates M-6701-1 and M-4 11-d M-4311-1 and.weld 101-142'as determined by impact testing, or

{2) the predicted shift in reference temperature for we]d an) plate

)Hi~ as determined by RG 1.99, "Effects of Residual Elements on II Predicted Radiation bamage to Reactor Vessel Materials.

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PLANT SYSTEMS 3/4. 7. 7 CONTROL ROOM ESSENTIAL FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION S l:VVi LL Aii.i"i7 IJ<Jf f 3.7.7 Two independent control room essential filtration systems shall be OPERABLE.

APPLICABILITY: Al 1 MODES.

ACTION:

MODES 1, 2, 3, and 4:

With one control room essential filtration system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MODES 5 and 6:

a ~

b.

With one control room essential filtration system inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE control room essential filtration system in the recirculation mode.

With both control room essential filtration systems inoperable, or with the OPERABLE control room essential filtration system, required to be in the recirculation mode by ACTION a., not capable of being powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREMENTS 4.7.7 Each control room essential filtration system shall be demonstrated OPERABLE:

At least once per 31 days on a

STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes.

P'.<

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber

housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

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PALO VERDE " UNIT 1 3/4 7"15 A88 gjgolg4-

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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued)

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C 1.

Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C. 5. a, C. 5. c and C. 5. d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 28,600 cfm + 10K.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

3.

Verifying a system flow rate of 28,600 cfm + 10K during system operation when tested in accordance with ANDRI N510-1980.

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After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal 'that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of 'Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

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At 1:east once per 18 months by:

l.

Verifying that the pressure drop acress the combined HEPA filters, pre-filters, and charcoal adsorber banks is less than 8.4 inches Water Gauge while operating the system at a flow rate of 28,600 cfm + 10K.

2.

Verifyirig that on a control room essential filtration actuation, the system is automatically placed into a filtration mode of operation with flow through the HEPA filters and charcoal adsorber banks.

3.

Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/4-inch Mater Gaug relative to during system operation CRT~

Q PALO VERDE - UNIT 1 3/4 7"16 4~8 ~/io/r+

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OCT 1 1984 MEETING NOTICE DISTRIBUTION:

NRC PARTICIPANTS ELicitra JDonohew bcc:

Applicant 8 Service List

~Doc e

~o(st 50-528 NRC PDR Local PDR TIC NSIC PRC System LB3 Reading H. Denton/E.

Case D. Eisenhut/F.

Miraglia T. Novak J.

Youngblood A. Schwencer E.

Adensam E. Butcher D. Crutchfield C. Grimes, G. Holahan C.

Thomas G. Lainas S.

Varga D. Vassallo J. Miller J, Stolz R. Vollmer W. Johnston J.

P. Knight R. Bernero L. Rubenstein W. Houston D. Muller T. Speis F. Schroeder H. Thompson W. T. Russell ACRS (16)

Attorney, OELD E. L. Jordan N. Grace F.

Ingram, PA Receptionist (Onlyjf. meeting is held in Bethesda)

Project Manager J.

Lee

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