ML18227C825

Letter Submitting Three Signed Originals and Forty Conformed Copies of Request to Amend Facility Operating License DPR-31 and DPR-41

ML18227C825
Person / Time
Site:	Turkey Point
Issue date:	06/21/1974
From:	Robert E. Uhrig Florida A & M University, Florida Power & Light Co
To:	O'Leary J US Atomic Energy Commission (AEC)
References
Download: ML18227C825 (64)

Text

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CONTROL NO:

FILE DATE OF DOC'ATE REC'D TWX RPT OTHER Florida Power & Light Company Miami, Florida 33101 Robert E. Uhrig 6-21-74 6-28-74 TO:

M O'Lea CLASS UNCLASS XXXX PROP INFO ORIG

~3ei eed INPUT CC OTHER 40 No C S REC'r SENT AEC PDR X

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50-2W 251 DESCP1PTXON:

Ltr notarized. 6-21-74, trans the following:

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Tee ts for Dist PLANT NAME: Turkey Point Units 3 & 4 ENCLOSURES:

Amdt to Facility Oper. Lic., consisting of changes to the Tech Specs for the Turkey 4k'IfgdggEg De:Nkl Htlrii@ffl FOR ACTION/INFO&IATIGN 7-1-74 AB BUTLER(L)

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John F. O'eary, Direc Directorate of Licensing Office of Regulation U.

S. Atomic Energy Commission Washington, D.

C.

20545

Dear Mr. O'eary:

Re:

Turkey Point e

Units 3 & 4 Docket Nos. '-250

& 50-251 Proposed Amen en to Facility 0 eratin Lic'enses DPR-31

& DPR-'41 FLORIDA POWER & LIGHT COMPANY June 21, 1974 42 P/Q(ypg Jtjl pa ca ca~

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In accordance

'with 10 CFR 50.59, Florida Power

& Light Company submits herewith three signed originals and forty (g0) conformed copies" of a proposed amendment to Facility Operating Licenses DPR-31

& DPR-41.

The changes are as set forth in the revised Technical Specifications'ages (Appendix A to DPR-31 and DPR-41) and are as described below:

P~ae 1-3 The definition of channel check has been modified to include source check of area and process radiation monitoring systems where comparison with another independent channel measuring the same variable is not practical.

~Pa e 1-5 Abnormal occurrence definitions (4) and (6) have been changed to clarify the definition and bring it into conformance, with Regula-tory Guide 1.16.

Subsections 1.17, 1.18, 1.19 and 1.20 have been added to define terms used in the Technical Specifications which were not previously defined.

HELPING SUILD FLORIDA

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Mr. John F.

O'L~~ry June 21, 1974

2. 1-1 Subsections 2.1, 2.1.1, and 2.1.2 have been changed to show that the average reactor coolant temperature is the limiting temperature.

Pa es 2.3-2 and 2..3'-3 The term f(hq) has been substituted for f(hI) to clarify the relationship between the specification and plant instrumentation and parameters.

Table'3'.'5'-2 The footnote to Item 1.4 has been changed to reflect operation at reduced RCS pressure.

Pa

'e's2 '3.'6'-'1'and'3'.'6-2 In subsection 3.6, specifications b,.3 and c.3 have been modified to clarify the intent, The plant has three boric acid tanks which are shared by the 'two, units.'he "modified wording clarifies the intent that adequate boric acid is available to both reactors.

~3; 14-2 In subsection 3.10, specification 5 has been changed to specify reactor coolant temper'ature because 'at low temperature with the Residual Heat Removal System in ser'vice there 'is no meaningful measurement of Tavg.

Coolant -temper'ature 'is measured by instrumentation in the Residual Heat Removal System.

Table 4'.'1'-1 A footnote has been added to Items 17A and 17B which eliminates the surveillance 'requirements for containment pressure when the equipment hatch or both doors of, either air lock are open.

~4'. '4'-'1 The applicability statement of subsection-4.4 has been expanded to include 'all the surveillance requirements of this subsection.

Subsection;4.,4.1 has been rewritten so that the integrated leak rate test will be performed and reported in compliance with 10 CFR 50, Appendix J.

Subsection 4.4.2 has been rewritten so that the local penetration tests will comply with 10 CFR 50, Appendix J, except that (1) the method of leak testing the air locks after each use has been changed.

This,ch'ange will facilitate the 'test without degrading

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June 21, 1974P the safety of the plant, and (2) the electrical penetration leak test requirement that the tests be performed during each refueling has been changed to require that the tests be performed prior to the integrated leak tests.

This is an exception to Appendix J only since Technical Specification 4.4.2 previously required these tests be performed prior to the integrated leak tests.

The previous subsection 4.4.3 has been deleted because the test reporting requirements are now contained in 10 CFR 50, Appendix J.

4; 4-3 Subsection 4.4.3 (previously designated subsection 4.4.4) has been rewritten so that the containment isolation valve tests will comply with 10 CFR 50, Appendix =J.

Subsection 4.4.5 has been redesignated subsection 4.4.4.

Specif-ication (f) has been changed to provide for a test of the Pesidual Heat Removal System each refueling.

This change eliminates an otherwise unnecessary shutdown and cooldown for both units.

Pa'o'e's '4.4-4,'4.4-'5'an'd. 4. 4-6 Subsection,4.4.6 has been redesignated subsection 4.4.5.

Subsection 4.4.7 has been redesignated subsection 4.4.6.

The dates of structural integrity testing of each containment have been added to clarify the requirements of the specification.

'0.4-7 Subsection 4..4.8 has been redesignated subsection 4.4.7.

A sentence was also added to the end of this subsection denoting the fact that liner surveillance has been satisfactorily completed,

~4. 6-1 Xn subsection 4.5, specification l.a has been changed to permit testing of the safety injection system by actually starting the residual neat removal pump.

This change eliminates the requirement that a residual heat removal pump be made inoperable during the test.

2~4'."6'-2 The acceptable level of performance for the emergency containment cooling fans has been clarified.

P~a'. 7'-1 Xn subsection 4.7, specification ->.7.1.1 a typographical error has been corrected.

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Mr. John,P.

O'eary June. 21, 1974 Pa es 4.8-1 and 4.8-2 In subsection 4.8, specification l.b,the second sentence has been changed.

This change

.was made 'to confor'm with the intent of the surveillance,'.e.,

that each 'diesel generator be thoroughly tested to ensure that the emergency power system will res'pond promptly and proper'ly.

In subsection 4.8, specification l.e,the first sentence has been changed to allow the 'requirements of subsection 4.8,specification l.a, to be exceeded during a test.

This is necessary to meet the requirement that the diesel gener'ator shall be loaded to 2750 kw.

If the load is not permitted below 2750 kw, then some overload must be allowed to provide 'for gover'nor swings.

The second sentence in specification l.e was also changed to clarify the intent of the specification.

In subsection,4.8, specifications 2.a and 2.b have been changed to reflect the addition of two new'tation batteries.

'4;10-1 In subsection 4.10, a statement was added to the end of specifi-cation 1 to preclude the 'requirement to test the auxiliary feed-water'umps during cold shutdown when testing is not possible.

The'erm f(hq) has replaced f(b,I) in the Reactor Coolant Temperature Bases.

This change was.made to be consistent with the proposed Technical Specifications.

These changes have been reviewed by the 'Plant Nuclear Safety Committee and the Company Nuclear Review Board.

These groups have independently determined that the changes'o not degrade but, in fact, improve 'the 'safety of the facility.

Very truly yours, o ert.E. Uhrig Vice President REU:nch Attach."

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Mr, Jack

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Newman

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STATE OF FLORIDA )

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COUNTY OF DADE

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Robert E. Uhrig, being first duly sworn, deposes and says:

That he is a Vice President of Florida Power 6 Li;ght Company, the Licensee her'ei'n; That he has executed the foregoing instrument; that the statements made in this said instrument are true and correct to the best of his knowledge, information and belief; and that he is authorized to execute the instrument of said

Licensee, Ro ert E.

rig Subscribed

.to and sworn to before me thi's' day of =

1974.

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ic xn an or t e of g de, St'ate of Florida uUrAIIYPUBLIC, SIAtE ot ILuttIOAaI lAktti:

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number of operable channels and the number of channels which when tripped will cause reactor trip.

1.7 INSTRUMENTATION SURVEILLANCE 1)

Channel Check Channel check is a qualitative determination of acceptable operability by observation of channel behavior during operation.

This determination shall include comparison of the channel with other independent channels measuring the same variable or radio-active source check of the Area and Process Radiation Monitoring Systems for channels.

2)

Channel Functional Test A channel functional test consists of infecting a simulated signal into the channel to verify that it is operable, in-cluding alarm and/or trip initiating action.

3)

Channel Calibration Channel calibration consists of the adjustment of channel output such that it responds, with acceptable range and ac-

curacy, to known values of the parameter which the channel measures.

Calibration shall encompass the entire channel, including alarm or trip, and shall be deemed to include the channel functional test.

1.8 SHUTDOWN 1)

Cold Shutdown The reactor is in the cold shutdown condition when the reactor is subcritical by at least 1% hk/k and T is less than 200F.

avg 2)

Hot Shutdown The reactor is in the hot shutdown condition when it 1-3 6/21/74

1. 13 ABNORMAL OCCURRENCE An abnormal occurrence is defined as any of the following:

l.

A safety system setting less conservative than the limiting set-ting established in the Technical Specifications, I

2.

.Violation of a limiting condition for operation established in the Technical Specifications.

3.

An uncontrolled or unplanned release of radioactive material from any plant system designed to act as a boundary for such material in an amount of significance with respect to limits prescribed in Technical Specifications.

4.

Incidents or conditions which prevented or could have prevented the performance of the intended safety function of an engineered t

safety feature system or of the reactor protection system.

5.

Abnormal degradation of one of the several boundaries designed to contain the radioactive materials resulting from the fission process'.

6.

Any uncontrolled or unanticipated change in reactivity equal to or greater than 1%.

K 7.

Observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the plant.

8.

.Conditions arising from natural or offsite manmade events that affect or threaten to affect the safe operation of the plant.

1.14 POWER TILT The power tilt is the ratio of the maximum to average of the upper out-of-core normalized detector currents or the lower out-of-core normalized detector currents whichever is greater.

If one out-of-core detector is out of service, the remaining three detectors are to be used to compute the average.

1-5 6/21/74

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1.15 INTERIM LIMITS" Limitations are imposed upon reactor core power and power distribution beyond previously established design bases consistent with interim bases for core cooling analysis established by the AEC in 1971 and bases for the effects of densification established in November 1972.

Interim power of the reactor core is limited to the values determined in accordance with specification 3.2.

Interim power in MWt equals N x 2200, where N is determined in accordance with Section 6.c. of specification 3.2.

The fuel residence time for cycle 1 shall be limited to 11,800 effective full power hours (EFPH) for Unit 3 and 24,500 EFPH for Unit 4 under low pressure operating conditions.

1 1.16 LOW POWER PHYSICS TESTS Low power physics tests are tests below a nominal 5% of rated power which measure fundamental characteristics of the reactor core and related instrumentation.

1.17 ENGINEERED SAFETY FEATURES Features such as containment, emergency core cooling, and containment atmospheric cleanup systems for mitigating the consequences of postulated il'ccidents.

1.18 REACTOR PROTECTION SYSTEM Systems provided to act, if needed, to avoid exceeding a safety limit in anticipated transients and to activate appropriate engineered safety features as necessary.

1.19 SAFETY RELATED SYSTEMS AND COMPONENTS

'hose plant features necessary to assure the integrity of the reactor coolant pressure

boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents which could result in off-site exposures comparable to'the guideline exposures of 10 CFR"100.

1.20 PER ANNUM During each calendar year.

1-6 6/21/74

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, average coolant temperature and flow during power operation.

~Ob ective:

To maintain fuel cladding integrity.

The combination of thermal power level, coolant pressure and average coolant temperature shall not exceed the limits shown in Figure 2.1-1 for full flow from three reactor coolant pumps.

2.

TWO LOOP OPERATION The combination of thermal power level, coolant pressure and average coolant temperature shall not exceed 1

the limits shown in Figure 2.1>>2 for full flow from two reactor coolant pumps'.

ONE LOOP OPERATION The thermal power level shall not exceed 20%,

coolant pressure shall be maintained in the 1820 2400 psig range, and the average coolant temperature shall not exceed 590 F for full flow from one reactor coolant pump.

4.

NATURAL CIRCULATION The thermal power level shall not exceed 12%,

coolant pressure shall be maintained in the 2135 2400 psig range and the average coolant temperature shall not exceed 602 F, when no reactor coolant pumps are in operation.

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Reactor Coolant Tem erature Overtemperature AT

< hT Kl 0.0174(T-566.6)

+ 0.000976(P-1885) - f(Aq) hT Indicated AT at rated

power, F

T

~ Average temperature, F

P Pressurizer

pressure, psig

'1 f(Aq) a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during

'tartup tests such that:

For (q - q ) within +10 percent and -14 percent t

b where q

and qb are the percent power in the top and bottom halves of the core respectively, and q

+ q is total core. power in percent of rated.

power, f(Aq) ~ O.

For each percent that the magnitude of (q q )

t b

exceeds

+10 percent, the Delta-T trip set po'int shall be automatically reduced by 3.5 percent of

'ts value at interim power.

For each percent'hat the magnitude of (q q )

t b

exceeds

-14 percent, the Delta-T trip set point shall be automatically'educed by 2 percent of.

its value at interim power.

Kl (Three Loop Operation)

~ 1;120 (Two Loop Operation)'

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Overpower hT

< AT

1. 09 K K (T T '

f (Aq) 1 dt 2

hT Indicated hT at rated power, F

0 T

Average temperature, F

T'ndicated average temperature at nominal conditions and rated power, F

K

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0 for decreasing average temperature, 1

0.2 sec./F for increasing average temperature K2 0.00134 for T equal to or more than T';

0 for T less than T'ate of change of temperature F/sec dT dt f(Aq) -

As'defined above Pressurizer Low Pressurizer pressure equal to or greater than 1715 psige High Pressurizer pressure equal to or less than 2385 psig.

High Pressurizer water level - equal to or less than 92% of full scale.

Reactor Coolant Flow Low reactor, coolant flow equal to or greater than 90% of normal indicated flow Low reactor coolant pump motor frequency equal to or greater than 56.1 Hz Under voltage on reactor coolant'pump motor bus equal to or greater than 60% of normal voltage Steam Generators Low-low steam generator water level equal to or greater than 5% of narrow range instrument scale 2 ~ 3 3

6/21/74

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TABLE 3.5-2 ENGINEERED SAFETY FEATURES ACTUATION NO.

FUNCTIONAL UNIT 1.

SAFETY INJECTION 1.1 Manual 1.2 High Containment Pressure 1.3 High Differential Pressure between any Steam Line and the Steam Line Header MIN.

OPERABLE CHANNELS

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MIN.

DEGREE OF REDUN-DANCY OPERATOR ACTION IF CONDITIONS OF COLUMN 1 OR 2 CANNOT BE MET Cold Shutdown Cold Shutdown Cold Shutdown 1.4 Pressurizer Low Pressure"and Low Level 1.5 High Steam Flow in 2/3 Steam Lines with Low Tavg or Low.

Steam Line Pressure 2**

1 1/line in 1

each of 2 lines Cold Shutdown Cold Shutdown 2.

CONTAINMENT SPRAY 2.1 High Containment Pressure and High-High Containment Pressure (Coincident) 2 per set 1/set Cold Shutdown This signal may be manually bypassed, when the reactor is shut down and pressure is below.18OO psig Each channel has two separate signals 6/21/74

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3.6.

CHEMICAL AND VOLUME CONTROL SYSTEM Volume Control System.

~Ob ective:

To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.

e flow path to the core for boron injection.

b, A reactor shall not be made critical unless the following Chemical and Volume Control System conditions are met:

1.

TWO associated charging pumps shall be operable.

2.

TWO boric acid transfer pumps shall be operable.

3.

The boric acid tanks in service shall contain a total of at least 3,080 gallons of a 20,000 to 22,500 ppm boron solution at a temperature of at least II 145 P.

4.

System piping, interlocks and valves shall be operable to the extent of establishing one flow path from the

, boric acid tanks, and one flow path from the refueling water storage tank, to the Reactor Coolant System.

5.

TWO channels of heat tracing shall be operable. for the flow path from the boric acid tanks.

The'rimary water storage tank contains not less than 30,000 gallons of water.

c.

The second reactor shall not be made critical unless the following conditions are met:

3,6~1, 6/21/74

1.

TWO associated charging. pumps shall be operable.

2.

THREE boric acid transfer pumps shall. be. operable.

3.

The boric acid tanks in service shall contain a total of at least 6160 gallons of a 20,000 to 22,500 ppm boron solution at a temperature, of at least 145 F.

4.

System piping, interlocks and valves shall be operable to the extent of establishing one flow path from the boric acid tanks, and one flow path from the refueling water storage tank, to each Reactor Coolant System.

5.

TWO channels of heat tracing shall be operable for the" flow path from the boric acid tanks.

6.

The primary water storage tank contains not less than 30,000 gallons of water.

d.

During power operation, the requirements of 3.6.b and c

may be modified. to allow one of the following components to be inoperable.

If the system is not restored to meet the requirements of 3.6b and c within the time period specified, the reactor(s) shall be placed in the hot shutdown con-dition. If the requirements of 3.6.b and c are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor(s) shall be placed in the cold shutdown condition.

l.

One of the two operable charging pumps may be removed from service provided that it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

One boric acid transfer pump may be out of service provided that it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

One channel of heat tracing may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.6-2 6/21/74

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5.

At least ONE residual heat removal pump shall be in operation, unless reactor coolant temperature is less than 160F.

6.

When the reactor vessel head is removed and fuel is, in the vessel, the minimum boron concentration of 1950 ppm shall be maintained in the reactor coolant system and verified daily.

7 Direct communication between the control room and the refueling cavity manipulator crane shall be available during refueling operation.

8, The spent fuel cask shall not be moved over spent fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit.

9, Fuel which has been discharged from a reactor will not be moved outside the containment in fewer than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.

If any one of the specified limiting conditions for re-"

fueling is not met, refueling shall cease until specified limits are met, and there shall be no operations which may increase reactivity.

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TABLE 4.1-1 (Continued)

Channel Descri tion Check Calibrate Test Remarks 10.

Rod Position Bank Counters ll.

Steam Generator Level 12.

Charging Flow N.A.

N.A.

N.A.

N.A.

With Analog Rod Position 13.

Residual Heat Removal Pump Flow 14.

boric Acid Tank Level N.A.

N.A.

N.A.

15.

kefueling Water Storage Tank Level 16.

Volume Control Tank Level 17A. Containment Pressure 17B. Containment Pressure 18A. Process Radiation 18B. Area Radiation 19.

Boric Acid Control 20.

Containment Sump Level 21.

Accumulator Level and Pressure N.A.tt tt W

N.A.

N.A.

St N.A.

R N.A.

N.A; tt tt M

N.A.

N.A.

Wide Range Narrow Range 22.

Steam Line Pressure

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TABLE 4. 1-1 (Continued)

Channel Descri tion Check Calibrate Test Remarks 23.

Environmental Radiological Monitors N.A.

A(1)

M(l)

(1) Flow 24.

Logic Channels N.A.

N.A.

25.

Emer. Portable Survey Instruments 26.

Seismograph N.A.

N.A.

N.A.

Using moveable in-core detector system.

Frequency only Effluent monitors only.

Calibration shall be as specified in 3.9.

Q Make trace Test battery (change semi-annually)

B/W M

, R A

Each Shift Daily Weekly Every Two Weeks Monthly Quarterly Prior to each startup if not done previous week Each Refueling Shutdown Annually Not applicable during cold or refueling shutdowns.

The specified tests, however, will be performed within one

'surveillance interval prior to startup.

N.A. when the equipment hatch or both doors of-either air lock are open.

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4.4 CONTAINMENT TESTS

~Ob ective:

Applies to containment integrity testing, tendon sur-veillance, end anchorage concrete surveillance, and liner surveillance.

To verify that potential leakage from the containment and the tendon loading are maintained within specified limits.

4.4.1 INTEGRATED LEAKAGE RATE TEST POST OPERATIONAL Post Qperational Containment Integrated Leak Rate Tests shall be performed and reported in accordance with 10 CFR 50, Appendix J, (type A tests).

Pa, the peak calculated containment internal pres-sure related to the design basis accident is 49.9 psig.

Pt, the containment vessel reduced test pressure is 25 psig.

La, the maximum allowable leakage rate at pressure Pa is 0.25 weight percent of containment atmosphere per day.

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4.

LOCAL PENETRATION TESTS Local penetration tests of the containment purge valves, the personnel and emergency air locks, the equipment access

opening, the fuel transfer tube flange, and the electrical pene-'rations shall be performed in accordance with 10 CPR 50 Appendix J, (type B tests) with the following exceptions:

1.

The personnel and emergency air locks shall be tested after each use by applying a 15 inch Hg minimum vacuum between the inner door gaskets.

2.

The electrical penetrations shall be tested prior to each containment integrated leak rate test.

4. 4-2 6/21/74

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4.4.3 ISOLATION VALVES Containment isolation valves shall be tested in accordance with 10 CFR 50, Appendix J, (type C

tests).

4.4.4 RESIDUAL HEAT REMOVAL SYSTEM a.

The portion of the Residual Heat Removal System that is downstream of the first isolation valve outside the containment shall be tested either by use in normal operation or hydrostatically tested at 350 psig at the interval specified below.

,b.

Visual inspection shall be made for excessive leakage from components of the system.

Any visual leakage

.that cannot be stopped at test conditions shall be measured by collection and weighing or by another equivalent method.

c.

The acceptance criterion is that maximum allowable leakage from the Residual Heat Removal Syst: em components (which" includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.

,d.

Repairs shall be made as required to maintain leakage with the acceptance criterion in (c) above.

e.

~ If repairs are not completed within 7 days, the reactor shall be shut down and depressurized until repairs are

'k effected and the acceptance criterion in (c) above is satisfied.

f.

Tests of tho Residual Heat Removal System shall be conducted each refueling.

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0 4.4.5

'TENDON SURVEILLANCE Lift-off Lift-offreadings will be taken for the following nine (9) tendons available for inspection:

Unit 3 Unit 4 Horizontal 62H18,42H70,64H50 13H15,51H50,35H70 Vertical 23V1,45V7,61V1 12V29,34V29,56V29 Dome 1D27, 2D28,3D28 1D28,2D28,3D28 Wire Ins ection One horizontal, one vertical and one dome tendon will be relaxed and one wire will be removed from each as a sample.

(At subsequent inspections different tendons will be used for the sample).

Wires will be visually inspected for corrosion and pitting.

Tensile tests will be performed on three (3) samples cut from each wire (one from each end and one from the middle) of a length equal to the maximum length acceptable for the test apparatus to be used.

After samples are. taken, tendons will be re-tensioned and final lift-offreadings will be taken.

Test Fre uenc Lift-offreadings and wire inspection will take place at the end of the first, third and every fifth year 'thereafter from the date of the structural integrity test (July 4, 1971, for Unit 3 and February 19, 1972, for Unit 4).

Tendon surveil-lance may be conducted during reactor operation.

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Additional Surveillance on Unit 3 Dome On Unit 3 dome 12 tendons (including the three listed in the first paragraph under 4.4.6) will be subjected to surveillance testing at 6, 12, 24 and 36 months after the structural integrity test (July 4, 1971, for Unit 3).

The ad-ditional tendons are:

1D15, 1D18,

1D36, 2D24, 2Dll,

2D21, 3D4, 3D21 and 3D24.

Lift-offreadings will be taken on each of the'se tendons.

The decrease in prestress force measured from 0.73f's A will be recorded and compared with the predicted loss, for the period the tendons were stressed.

The surveillance tendons will be stressed to 0.8f's, and the elongation recorded, the tendons will then be relaxed and observation will be made at the stressing washer for any indications of a broken wire.

The tendons will be re-tensioned and lift-offreadings taken.

Wire Ins ection One wire'ach will be removed from three

tendons, not listed in the first paragraph. of 4.4,6 (one from each directional group); wires will be visually inspected for corrosion and pitting.

Tensile tests will be performed on three (3) samples cut from each wire (one from each end and one from the middle) of a length equal to the maximum length acceptable for the test apparatus to be used.

After the samples are taken,'he tendons will be retensioned and final lift-offreadings taken.

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4.4. 6 END ANCHORAGE CONCRETE SURVEILLANCE The following end anchorages will be sub)ect to surveillance at the 346'uttress on Unit 3 and the 194'uttress on Unit 4:

35r 0

60>

0<<

85'r>

110'>>0", 152'-0" and in the tendon inspection gallery of each unit at tendon numbers:

12Vl1 12V23 23V9 23V23 ) 34V12i 34V28 45V14 j 45V26, 46V24, 56V16, 61V9, 61V26.

I The inspection intervals will be approximately one-half year and one year after the structural integrity test (July 4, 1971, for Unit 3 and Feb-ruary 19, 1972, for Unit 4) and shall be chosen such that the inspection occurs during the warmest and coldest part of the year following the test.

The inspections made shall include:

l.

Visual inspection of the end anchorage con-crete exterior surfaces.

2.

The mapping of the predominant visible con-crete crack patterns.

3.

The measurement of the crack widths, by use of optical comparators or feeler gauges.

h The measurements and observations shall be com-pared with those to which prestressed structures have been subjected in normal and abnormal load conditions and with those of preceding measurements and observations at the same location on the structures

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If the inspections determine that the conditions are favorable in comparison with experience and predictions, the close inspections will be terminated by the second inspection and a report will be prepared; If the inspections detect symptoms of greater than normal cracking or movements, an investigation will be made to determine the cause.

4.4.7 LINER SURVEILLANCE Three representative areas of the liner plate shall be examined and measured for inward deformation (1) prior to the structural in-tegrity test, (2) after the structural integrity test (3) approximately one year after the structural integrity test.

Measurements shall be taken between vertical anchors using a

straight edge to determine liner profile to within a + 0.01 inch accuracy.

If changes are less than 0.25 inches no further tests or action is required other than preparation of records.

Otherwise an investigation and cor-rective action will be taken.

Measurements locations shall be:

Elevation Unit 3 Unit 4 34 l 011 62'0" 118 l Oil 70'90'26'0'90'18'hen measurements are made, liner plate and ex-terior concrete surface temperatures in the area of measurement, and inside and outside ambient temp-eratures, will be determined and recorded.

The requirements of this Technical Specification have been met.

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4.5 SAFETY INJECTION Applies to testing of the Safety Injection System.

~Ob ective:

To verify that the sub)ect systems will respond promptly and perform their design functions.

S ecifications:

1.

SYSTEM TESTS a.

System tests shall be performed at each refueling shutdown.

The test shall be performed in accordance with the follow-ing procedure:

With the Reactor Coolant System pressure equal to or less than 350 psig and temperature equal to or less than 350F, a test safety infection signal will be applied to initiate operation of A the system.

The breakers for the residual heat

. removal pump motors will be tested either in the test position or by actually'perating the associated residual heat removal pump motor.

b.

The test will be considered satisfactory if con-trol panel indication and visual observations indicate that all components have received the safety injection signal in the proper sequence and timing, appropriate breakers shall open and

close, and all automatic valves shall complete their travel.

2.

COMPONENT TESTS e.

~Fum e

1.

The safety in)ection pumps and residual heat removal pumps shall be started at intervals not greater than one month.

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2.

COMPONENT TESTS Pum s and Fans The containment spray pumps. and the Emergency Coritain-ment Cooling fans shall be started at intervals not greater than one (1) month.

Acceptable levels of performance shall be that the pumps reach their rated shut off heads, the fan motors reach

their nominal operating current for the containment at-mosphere during the test, and both operate for at least fifteen minutes'.

Valves The systems motor operated isolation valves will be tested for operation during system tests.

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4.7 EMERGENCY CONTAINMENT FILTERING AND POST ACCIDENT CONTAINMENT VENT SYSTEMS Applies to the Emergency Containment Filtering and the Post Accident Containment Vent System components.

To verify that these systems and components will be able to perform their design functions.

EMERGENCY CONTAINMENT FILTERING SYSTEM l.

OPERATING TESTS System tests shall be performed at approximately quarterly intervals.

These tests shall consist of visual inspection and pressure drop measurements across each filter bank.

Visual inspection shall include.inspection of general condition for evidence of:

water, oil, or other foreign material; gasket deterioration; adhesive deterioration in the HEPA units; exces-sive dust cake on the demisters; and unusual or ex-cessive noise or vibration when the fan motor is A

running.

Pressure drop across any filter bank shall not exceed two times the pressure drop when new and shall not be less than the pressure drop when new..

2.

PERFORMANCE TESTS During each refueling operation, "in-place" DOP and freon tests shall.be conducted at design flow on each unit (all flow paths).

99.9%

'DOP removal and 99.5% freon removal shall constitute acceptable performance.

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4.8 EMERGENCY POWER SYSTEM PERIODIC TESTS ments of the emergency power system.

~Ob ective:

To verify that the emergency power system will respond promptly and properly.

as stated:

1.

DIESEL GENERATORS a.

Each diesel generator shall be manually started and synchronized with normal power sources and loaded to 2750 KW monthly.

b.

Each'diesel generator shall be started auto-matically by a simulated loss of all normal A-C power supplies together with a simulated safety injection signal and loaded sequentially with vital loads during each refueling shutdown.

Each diesel shall start and assume loads in the time sequence stated in FSAR Table 8.2-3.

The safety injection pumps will be operated using the test lines.

c.

Each diesel generator shall be given a thorough inspection at least annually following the man-ufacturer's rec'ommendations for this class of stand-by service.

d.

The above tests will be considered satisfactory if all applicable equipment operates as designed.

e.

Diesel generator electric loads shall not be in-creased beyond 2850 KW during a test.

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connected loads shall not be increased above those listed in FSAR Table 8.2-2 during the test in l.b. above.

f.

The diesel fuel oil transfer pumps shall be tested monthly.

2.

STATION BATTERIES a.

Pilot cell specific gravities. shall'.be".read and recorded daily.

The pilot cell shall be rotated on a monthly basis'.

Monthly each battery shall be'.,given",an equalizing charge, and afterwards specific gravity and voltage readings shall be taken and recorded for each cell.

Mater shall be added to restore normal level and total water use shall be recorded.

Complete visual in-spection of batteries shall be made monthly.

c.

Quarterly detailed visual inspection shall be made of chargers.

d.

Annually connections shall be checked for tightness and 'anti-corrosion coating shall be applied to interconnections.

e.

Perform load test annually.

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AUXlLlARYFEEDWATER SYSTEM Applies, to periodic testing requirements of the auxiliary feedwater system.

~Ob ective:

To verify the operability of the auxiliary feedwater system and its ability to respond properly when required.

S ecifications:

l.

Each turbine-driven auxiliary feedwater pump shall be started at intervals not greater than one month, run for 15 minutes and a flow rate of 600 gpm es-tablished to the steam generators.

The monthly frequency is not intended to require the test while at cold shutdown.

The testing requirement is met by performing this test during startup subsequent to cold shutdown.

2.

The auxiliary feedwater discharge valves shall be tested by operator action during pump tests.

3.

Steam supply and turbine pressure valves shall be tested during pump tests.

4.

These tests shall be considered satisfactory if control panel indication and visual observation of the equipment demonstrate that all components have operated properly.

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The f(hq) function in the Overpower hT and Overtemperature 5T protection system setpoints includes effects of fuel densification on core safety limits.

The setpoints will ensure that the safety limit of centerline fuel melt will not be reached and DNBR of 1.30 will not be violated.

(10)

Pressurizer The low pressuiizer pressure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident. (6)

The high pressurizer pressure reactor trip is set below the set pressure of the pressgrizer safety val'ves and limits the reactor operating pressure range.

The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief.

The specified set point allows margin (3) for instrument error and transient level overshoot before the reactor trips'eactor Coolant Flow The low flow reactor trip protects the core against DNB in the event of loss of one or mqre reactor coolant pumps.

The set point specified is consistent with the value used in the ac-(7) cident analysis.

The low frequency and under voltage reactor trips protect against a decrease in flow.

The specified set points assure a react'or trip signal before the low flow trip point is reached.

The underfrequency trip set point preserves the coastdown energy of the reactor coolant pumps, in case of a system frequency decrease, so DNB does not occur.

The undervoltage trip set point will cause

'a trip before the peak motor torque falls below 100% of rated torque.

Steam Generators The low-low steam generator water, level reactor trip assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting of the auxiliary feed-(8)

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water system.

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