ML20009H240

Amend 41 to License DPR-72,authorizing Facility Power Level Increase & Correcting Typographical Error on Tech Spec Page 3/4 3-8

ML20009H240
Person / Time
Site:	Crystal River
Issue date:	07/21/1981
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20009H241	List: ML20009H240 ML20009H242 ML20009H245
References
TAC-11009, TAC-46962, NUDOCS 8108070088
Download: ML20009H240 (27)

Text

g ae.

,/

UNITED STATES NUCLEAR REGULATORY COMMISSION n

l WASHWGTO N. D. C. 20555

%,...../

FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNE1.L CITY OF GAINESVILLE CITY OF KISSlMMEE-CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES TOWISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING UTILITIES COMMISSION SEMIN0LE ELECTR*C COOPERATIVE, INC.

~

l CITY OF TALLAHASSEE DOCKET NO. 50-302 l

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT

' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 41 I

License No. DPR-72 1.

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

I A.

The applications for amendment by Florida Power Corporation, et al l

(the licensees) dated November 29,1978, February 28, 1979, j

November 20, 1979, and July 9,1981, and supplemental filings comply l

with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of-l the Comission's regulations and all' applicable requirements have been l

satisfied.

l e

I i

!l i

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(1) and 2.C.(2) of Facility Operating License No. DPR-72 are hereby amended to read as follows:

(1) Maximum Power Level Florida Power Corporation is authorized to operate the facility at a steady state reactor core power level not in excess of 2544 Megawatts (100 percent of rated core power level).

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 41, are hereby incorporated in the license.

Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.

l 3.

This license amendment is effective as of the date of its l

issuance.

t COR THE NUCLEAR REGULATORY COMMISSI0ft a

f 9

arrell G. Eisenhut, Director Division of Licensing Office of fluclear Reactor Regulation l

l

Attachment:

l Changes to the Technic.;l Specifications l

Date of Issuance:

July 21, 1981 I

e

3

' ATTACHENT TO LICENSE AMENDMENT NO. 41 l

'1 FACILITY OPERATING-LICENSE-NO. DPR-72 DOCKET NO. '50-302 Replace the following pages of the Appendix."A'_ Technical Specifications I

.with the enclosed pages'. The revised pages are identified by Amendment-number and contain vertical lines indicating the area of-change. The

. corresponding overleaf'pages are also provided to' maintain document completeness.

Renove Pages -

-Insert Pages 1-1 1-1 2-2 2-2 2-3 2-3 2-5

.2-5 2-6 6 2-7 2-7 l'-

B2-3 B2-3 l

l B2-4 B2-4 82-5 B2-5 l

B2-6 B2-6 il'

.B2-7 B2-7 l.

B2-3 B2-8' 3/4 2-3 3/4 2-3 3/4.3-2 3/4 3-2 3/43-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-3 3/4 3-8 l

l l'

j l

---

. ~.,_.

^.

t i

1.0 DEFINITIONS i

DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

THERMAL F0WER 1.2 THERMAL POWER-shall be the total reactor core heat transfer rate to the reactor coolant.

l RATED THERMAL. POWER'

\\

l 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2544 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combina-tion of core reactivity condition, pow (e level and average reactor-coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6. A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY 'vhen it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary l

equipment, that are required for the system, subsystem, train, ccaponent or device to perform its function (s), are also capable of performing their related support function (s).

CRYSTAL RIVER - UNIT ?

1-1 Amendment No. 41 e

..m.-

I

-DEFINITIONS REPORTABLE OCCURRENCE 1.7 A _ REP 03 TABLE OCCURRENCE shall be any of those ' conditions specified as a reportable occurrence in Revision 4 of Regulatory Guide 1.16, L

" Reporting of Operating Information - Appendix "A" Technical Specifications."

l CONTAll: MENT INTEGRITY, l

1.8 CONTAlfiMENT_ INTEGRITY shall exist when:

l a.

All penetrations required to be closed during accident con-ditions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation system, or 2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.

b.

All equipment hatches are closed and sealed, c.

Each airlock is OPERABLE pursuant to Specification 3.6.1.3, d.

The containment leakage ratec are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE.

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that.it responds with necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel.-including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

CHANNEL CALIBRATION may be performed by any series of sequential, over-iapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CRYSTAL RIVER - UNIT 3 1-2 l

m.

-er

---e

i 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2,.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of the reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in Figure 2.1 -1.

APPLICABILITY : MODES 1 and 2.

ACTION-Whenever" the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STANDBY within one hour.

4 REACTOR CORE 2.1.2 The combination of reactor THERMAL ~ POWER and AXIAL POWER IMBALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of three and fcur reactor coolant pump operation.

APPLICABILITY : MODE 1.

ACTION:

l Whenever the point defined by the combination of Reactor Coolant System flow, A7.IAL. POWER IMSALANCE and THERMAL POWER has exceeded the appropriate safety limit, be in HOT STAND 3Y within one hour.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor ' Coolant System pressure shall not exceed 2750 psig.

APPLICABILITY : MODES 1, 2, 3, 4 and 5.

ACT' ION :

MODES 1 and 2 Whenever the Reactor Coolant System pressure has ex-ceeded 2750 psig, be in HOT STANDBY with the Reactor l

Coolant System pressure within its limit within one hour.

MODES 3, 4

- Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

CRYSTAL RIVER - UNIT 3 2 -1 Amendment No. 17

2400 RCS PRESS'JRE H13H TRIP RC QUILET TEY?

g 2200 HIGH TRI?

.i ACCE?T ABL E 5

OPERATION

.u a

e 5

4 E

2000 O

SAFETY Lly.li 3

g!

54 E

UNACCE??A3LE C?!RATI2.N 1300

• ES PEEE3UEE LOW TRIP i

550 500 520 540 t

i Rea:::t Outle: Tec:erature. F 4

i 1

i i

i 1 m.e.p 7.1 1 v..

REACTCR CORE SAFETY LIMIT i

4 CRYSTAL RI'IER UtlIT 3 2-2 Amendment No. M, 41 i

w-

l 6

120 I

i

(- 29.12,112.0) 1

( + 16.8.112.C)

(

f P

,3g l

(-37.4.106.0) t ACCEPTASLE 4 ggg l

- PUMP CPERATION

. s (+ 3 4.C,1 C C.C)

I i

l 1

i

(-29.12,39.71)I 90

(+16.a,39.71) l l

(-3 7.4,33.71) ([

l l

t l

I 80 ACCE.2 TABLE 3S 4 PuvP CPERATION h,)(+ 34 c,77,73 }

j l

l l

l I

I 70 1

i l

1

=

60 t

6 o

1 1

p 50 l

j 2

i x

I w

l

~

40 l0 l

1 i

a I

l l

I I

30 i8 l

N l

t

_ 20 i

l 2_

I i

I i

e I

i l

o I

I O

10 0

l l

z I

l i

-60

-50

-40

-30

-20

-10 0

+10

+ 20

-30

+40

-50

-60 AXI AL 8CwER 'MSALANCE, %

i FIGURE 2.1 - 2 i

i' RE ACTOR CORE S A.:ETY LIMIT 1

l CRYSTAL RIVER - UNIT 3 2-3 Arnendment No..Mr, R, R 41 i

l

. ~

' I lSAFETYLIMITSANDLIMIiTNGSAFETYSYSTEMSETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PD.0TECTICN SYSTEM SET 70!NTS 2.2.1 The Reacter Protection System instrumentation set:cints shall be set censistent with the Trip Setpoint values shown -in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

i

~

With a Reacter Protection System instrumentatien setpoint less conserv-ative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel incperable and apply the applicable ACTION statement recuirement of Specification 3.3.1.1 until the channel is restored to GPERABLE status with its trip setpoint adjusted censistent with the Trip Setpoint value.

3 i

1 l

CRYSTAL RIVER - UNIT 3 2,4 y

m-n m

-e

- +

9 C;

y TABLE 2.2-1 5

REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS 9

FUNCTIONAL UNIT TRIP SETPOINT ALLOWADLE VALUES Q

1.

Manual Reactor Trip Not Applicable Not Applicable w

2.

Nuclear Overpower

< 104.88% of RATED TilERHAL POiTR

< 104.88% of RATED TilERMAL POWER With four pumps operating with four pumps operating

< 79.92% of RATED TilERMAL POWER

< 79.92% of RATED TilERMAL POWER With three pumps operating With three pumps operating 3.

RCS Outlet Temperature-liigh

< 618*F

< 618'F 4.

Nuclear Overpower Trip Setpoint not to Allowable Values not to exceed liased on RCS Flow and exceed the limit line of the limit line of Figure 2.2-1 7

AXIAL POWER IMilALANCEIII figure 2.2-1 m

5.

RCS Pressure-Lowill 1 1800 psig

> 1800 psig k

6.

RCS Pressure-liigh

< 2300 psig -

< 2300 psig 7.

RCS Pressure-Variable tow (1) 1 (11.59 Tout F - 5037.8) psig

> (11.59 Tout "F - 5037.8) psig 5

O N

N e

S5#

TABLE 2.2-1 (Continued) y REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS 9

[

FUNCTION UNIT TRIP SETPOINT ALLOWADLE VALUES

c U

8.

Nuclear Overpower Based on More than one pump not operating More than one pump not operating w

Reactor olant Pump Power Monitors 9.

Reactor Contaimnent Vessel Pressure liigh

< 4 psig i 4 psig l

1 l

'?

a, III rip may be manually bypassed when RCS pressure < 1720 psig by actuating Shutdown Bypass provided T

that:

g co E

a.

The Nuclear Overpower Trip Setpoint is < 51 of RATED TilERMAL POWER 2

b.

The Shutdown Dypass RCS Pressure - liigh Trip Setpoint of < 1720 psig is imposed, and 5

c.

The Shutdown Hypass is removed when RCS Pressure > 1800 psig.

E.

E 1

o WW6

ll s,

4 t

12t,'

i I

I

{

I i

l 110

(.11.0,107.0) r

- (+ 0.0,10 7.0 )

l m.o,.73Q/

' '<2-- o 819 i l

l 8

(-26.0,9 6.0) l

}

ACCEPTABLE 4 90 PUMP OPERATION ft 22.0,89.0) l l

l

(-11.0,7 9.92)

- (0.0,79.92) l i

l I

i i

70 i

i i.26.0,6832)(

ACCEPTA8LE 384 l

l l

1 PUMP OPERATION s

60

)(+ 22.0,6a.9N; I

e i

l l

5 i

l I

l o

50 l

a Y

i I

g I

i 40 I

l C

l i

l 0

l I

2 30 i

I

=

l g

l I

20 i

l l

{

l 5

i to O

d I

i

=

i 1

-60

-50

-40

-30

-20

-10 0

+10

+20

+30

+40

+50

+60 AXt AL power IMS ALANCE, %

l FIGUR.

2.2-1 TRIP SETPolNT FOR NUCLEAR OVERPOWER 2ASED ON RCS FLOW AND AXIAL POWER IMBALANCE CRYSTAL RIVER - UNIT 3 2-7 Amendment flo, y, X 41

~

.=.-.

i SAFETY LIMITS i-BASES For each curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22% for that particular reactor coolant pump situation. The 1.30 DNBR curve for three pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of the three pump ~ curve will be above and to the left of the other curves.

2.1.3 REACTOR COOLANT SYSTEM PRESSURE' The restriction of 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 pressure vessel' and pressurizer are designed to Section l

III of the ASME Boile. and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor-Coolant Systen piping, valves and fittings, are designed to USAS B 31.7, February,1968 Draf t Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2760 psig is therefore consistent with the design criteria and associated code requirements.

l The entire Reactor Coolant System is hydrotested at 3125 psig,125%

of design pressure, to demonstrate integrity prior to initial operation.

i l

l l

CRYSTAL RIVER - UNIT 3 8 2-3 Amendment No.,L8', 41 1

\\

r 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set?oint 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 Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures.

The purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated.

The Nuclear Overpower Trip Setpoint of < 5.0t prevents any significant reactor power from being produced.

Sufficient natural circulation would be availaole to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentation channels and provides manual reactor trip capability.

Nuclear Overpower A Nuclear Overpower trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip i.s initiated when the reactor power level reaches 104.88* of rated power.

Due to calibration and instrument errors, the maximum actual power it which a trip would be actuated could be 112%, which was used in the safety analysis.

CRYSTAL RIVER - UNIT 3 8 2-4 Amendment No. M, 41

LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temperature - High The RCS Outlet Temperature High trip < 618'F prevents the reactor outlet temperature from exceeding the designlimits and acts as a backup trip for all power excursion transients.

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE Tne power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has bet established to ac-commodate flow decreasing transients from high power.

The power level trip setpoint produced by the power-to-flow ratio pro-vides both high power level and low flow protection in the event the re-actor power level increases or the reactor coolant flow rate decreases.

The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every pow-er level there is a minimum permissible low flow rate.

Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if power is > 107% and reactor flow rate is 100%, or flow rate is 193.45% and power level is 100t'.,

2.

Trip would occur when three reactor coolant pumps ' ' operat-ing if power is > 79.92% and reactor flow rate is. +.7%, or flow rate is 176.09% and power is 75%.

For safety calculations the maximum calibration and instrumentation er-rors for the power level were used.

CRYSTAL RIVER - UNIT 3 8 2-5 Amendme t No. Rr, F, X,41

i LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMS/J.ANCE bcundaries are established order te prevent reactor thermal limits from being exceeded. These the '.' limits are either power peaking kw/f t limits or DNBR limits. The rXIAL POWER IM-BALANCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced. The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.077, for a li, flow reduction.

RCS Pressure - Low, High, and Variable Low The High and ' Low trips are provided to limit the pressure range in which reactor operation is permittd.

During a slow reactivity insertion startup accident from lcw power or a slow reactivity insertion from high power, the RCS Pressure-High set-point is reached before the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure-High, 2300 psig, has been established to main-tain the system pressure below the safety limit, 2750 psig, for any de-sign transient. The RCS Pressure-High trip is backed up by the pressur-izer code safety valves for RCS over pressure protection and is, there-fore, set lower than the,et pressure for these valves, 2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip.

The RCS Pressure-Low,1800 psig, and RCS Pressure-Variable Low, (11.59 Tout F - 5037.8) psig, Trip Setpoints have been established to maintain the DNS ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction.

It also prevents reactor operation at pressures below the valid range of DN3 correlation limits, protecting against DNS.

Due to the calibration and instrumentation errors, the safety analysis used a RCS Prenure-Variable Lew Trip Setpoint of (11.59 Tout

• F - P137.8) psig.

CRYSTAL RIVER - UNIT 3 8 2-6 Amendment No. M, X, E, 41 l

l tIMITING SAFETY SYSTEM SETTINGS BASES Reacter Containment Vessel Pressure - Hich The ' Reactor Containment Vessel Pressure-Hign Trip Set;cint < 4

sig, provides positive assurance that a reactor trip will occur ~in the unlikely event of a steam line failure in the r.cntainment vessel or a less-of-coolant accicent, even in the absence of a RCS Pressure - Low t ri p.

Reactor C'oclant Pump Power Monitors In conjunction with the power / imbalance / flow trips, the Reactor Coolant Pump Power Monitors trip prevents the minimum core DUBR from decreasing below 1.30 by tripping the reactor due to more than one reactor coolant pump /not operating.

A reactor coolant pump is considered to be not operating when the power required by the pump is >l20% or is <707. of the nominal operating power. The nominal operating power decreases from when a pump is first started during heatup and is pumping dense fluid (typically 7500KW) to when a pump is,oerating at full reactor power and is pumping less dense fluid (typically 5500KW).

In order to avoid spurious trips during normal operation, the 120% trip setpoint (9000KW) is based on the nominal operating power for a pump during heatup and the 70% trip set-point (3900KW) is based on the nominal operating power for a pump operating at full reactor power.

CRYSTAL RIVER - UNIT 3 3 2-7 Amendment No..Hr,.W,41 i

I('

r I

I 2400 i

i t

I i

l l

3 CURVE 2 i

.3 PUMP i

2200-f i

i i

I I

5 kJz l

D 4

I e

I M

l s

CURVE 1 W

j

[

l j

4 PUMP a

4 w

a i

i l

i i

t 2000 L

i l

I l

w i

e i

l o

1 t

v 3

l f

4 6

n 5

i g

1800 I

[

I i

l i

I l

580 600 620 640 l

REACTOR OUTLET TEMPER ATURE, F REACTOR COOL ANT FLOW PUMPS OPERATING CU VE.

F LOW (% D ESIG N1 P OWER (RTP)

(TYPE OF LIMIT)

I 139.7 x 106 (10 6.5 % )

113.0 5 %

4 PUMPS (DNSR) 2 104.4 x 10 s ( 79.6 %)

90.84 %

3 PUMPS (DNSR)

PRESSURE / TEMPER ATURE LIMITS AT M AXIMUM ALLOWABLE POWER FOR MINIMUM DNER B A SES FIGURE 2.1 CRYSTAL RIVER UNIT 3 B2-8 Amendment No. E, X,41

~~

t

+

l 6

1-

?

k 6

2 4

j-i i

i

~

i THIS FIGURE IS DELETED FOR THE REMAINDER OF CYCLE 3 4

1 I'

i 1

i

.i 4

i 1

4 4

i

'l 1

1

].

I i

4 1

4-1 FIGURE 3.2-2 1i i

i i

e i

t

} -

CRYSTAL RIVER - UNIT 3 3/42-3 1

Amendment No. E, X, 2 41 e

j.

>OO

DOWER DISTRIBUTION LIMITS NUC:. EAR HEAT FLUX HOT CH*';NEL FACTOR - F A

' IMITING CONDITICN FOR OPERATICN

3.2.2 Fqsnallbelimitedbythefollowingrelaticnships

F 3.08 q1 THEVAL POWER where P = RA ED B ERFAL PO'WER and P 1 1.0.

APoLICABILITY: MCDE 1.

ACTION:

With F cxceeding its limit:

g a.

Reduce THERFAL POWER at least 15 for eech 1% F exdeeds the limit within 15 minutes and similarly reduce t.9e Nuclear Overpower Trip Set;oint and Nuclear Overpcwer based en RCS Flew and AXIAL POWER IMBALAriCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Demenstrate through in-core mapping that F is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit $r reduce THER'tAL POWER to less than 5% of RATED THEP?AL POWER witnin the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c.

Identify and correct tne cause of the out of limit condition prier to increasing THE??AL PCWER above the reduced limit re-quired by a or b, above; subsequent POWER CPERAT!*.'i may proceed providr.d that F is demonstrated through in-core mac-ping to be v4*nin its 19mit at a nominal 50% of RATED THEFFAL F0VER price to exceeding this THEFPAL POWER, at a nominal 75%

of RATED THERMAL POWER prict to exceeding this THERMAL POWE.R and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERFAL POWER.

SURVEILLANCE REOUIREMENTS a.2.2.1 F shall be detemined to be within its limit by using the incere g

detectors.o cotain a pcwer distribution map:

CRYSTAL RIVER - UttIT 3 3/4 2-a Amendment No. 19.

3/4.3 INSTRUMENTATION l.

3/4.3.1 REACTOR-PROTECTION SYSTEM INSTRUMENTATION LIMITING' CONDITION FOR OPERATION 3.' 3.1.1 'As a minimum, the Reactor Protection System instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE

. TIMES as shown in Table ~3.3-2.

. APPLICABILITY: As'shown in Table 3.3-1.

' ACTION:

-As shown in Table 3.3-1.

I l

I SURVEILLANCE REQUIREMENTS

4. 3.1.1.1 Each Reactor Protection System instrumentation channel shall be demonstrated GPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIO"AL TEST operations during the MODES and at the frequencies shown in Table 4.3-1.

4. 3.1.1. 2 The total bypass function shall be demonstrated OPERABLE at L

least once per 18 months during CHANNEL CALIBRATION testing of each l

channel affected by bypass operation.

4.3.1.1.3 TheREACTORPROTECTiONSYSTEMRESPONSETIMEofeachreactor

' trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least or.ce every N times 18 months where N is the total number of redundant channels in a specific reactor

[

trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

l l

t CRYSTAL RIVER - UNIT 3 3/4 3-1 n

- - y

'\\

TABLE 3.3-1 REACTOR PROTECTI0li SYSTEM INSTRUMENTATION 95 MINIMUM TOTAL NO.

CHANiiELS-CHANNELS APPLICABf.E g

--OF CilANNELS TO TRIP OPERABLE H0 DES ACTION F

FUNCTI0fiAL UtilT 1.

Manual Reactor Trip 1

1 1

1, 2 and *'

8 g

2.

fluclear Overpower 4

2 3

1, 2 2#

E 3.

RCS Outlet Temperature--liigh 4

2 3

1, 2 '.

3f l

4.

fluclear Overpower Based on RCS Flow and AXIAL-POWER IMBALAflCE 4

2(a) 3 1, 2 2f 5.

RCS Pressure--Low 4

2(a)

.3 1, 2 3#

6.

RCS Pressu.c--liigh -

4 2

3 1, 2 3#

7.

Variable Low RCS Pressure 4

2(a) 3 1, 2 3#

8.

Reactor Containment Pressure--Iligh 4

2 3

1, 2 3f 9.

Intermediate Range,fleutron Flux 2

0 2

1, 2 and

• 4 and Rate w

10. Source Range, Neutron Flux and Rate A.

Startup 2

0 2

2ff and'*

5 B.

Shutdown 2

0 1

3, 4.and 5 6

11. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per I, 2 and

• 7f system system trip system F

12. Reactor Trip Module 2 per trip 1 per trip 2 per 1, 2 and

• 7#

system system trip system-o.

2

13. Shutdown Bypass RCS Pressure-liigh 4

'2 3

2**,

3**,

6#

5 4**, 5**

=

14. Reactor Coolant Pump Power Monitors 4

2(a) 3 1, 2 3#_

l.

f:

S' e

99we'

c l'

TABLE 3.3-1 (Continued)

ACTI0ft STATEMENTS (Continued) a.

< 5% of RATED THERIGL POWER restore the inoperable channel to OPERABLE status prior to increasing THERIGL POWER above 5% of RATED THERiGL POWER.

b.

> 5'; of RATED THER!GL POWER, POWER OPERATION may continue.

With the number of channels OPERABLE one less than required ACTION 5 by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. ' < 10- N amps on the Intermediate Range (IR) in-strumentation, restore the inoperable channel to OPE.1ABLE gatus prior to increadng THERMAL POWER above 10-amps on the IR instrumentation.

b.

> 10-10 amps on the IR instrumentation, operation may continua.

With the number of channels OPERABLE one less than re-ACTI0fi 6 quired by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 within one hour 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.

With the number of OPERABLE channels one less than the ACTION 7 Total Number of Channels STARTUP and/or POWER OPERATION ray prtceed provided all of the following conditions are satisfied:

a.

Within I hour:

1.

Place the inoperable channel in the tripped condition, or 2.

Remove power supplied to the control rod trip device associated with the inoperative channel.

b.

One additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, and the inoperable channel above may be bypassed for up to 30 minutes in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period when necessary to test the trip breaker associated with the logic of the channel being tested per Specification 4.3.1.1.

The inoperable channel above may not be bypassed to test the logic of a channel of the trip system associated with the inoperable channel.

With the number of channels OPERABLE less than required ACTIO'! 8 by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

CRYSTAL RIVER - UNIT 3 3/4 3-5

1.

TABLE 3.3-2 O

REACTOR PROT _E_CTION SYSTEM INSTRUMENTATION RESPONSE TIMES C

r-

-RESPONSE TIMES 5

FUNCT10NAL UNIT M

Not Applicable' 1.

Manual Reactor Trip

< 0.3' seconds E

2.

Nuclear Overpower

• G Not Applicable 3.

RCS Outlet Tenperature-liigh 4.

Nuclear Overpower Based on RCS Flow 1 1.4 seconds and AXIAL POWEll IMBALANCE

• I 0.5 seconds S.

RCS Pressure -- Low 1 0.5 seccnds.

M 6.

RCS Pressure -- liigh u-

^

.Not Applicable

{

7.

Variable Low RCS Pressure Not Applicable 8.

Reactor Contai nnent Pressure -- liigh 1 0.62 seconds 9.

Iteactor Coolant Pump Power Monitors S*

8

$r E

~

itesponse time of the neutron flux signal portion of Tii6dtrJn cie~tectors are exonal fran response time testing, the channel shalI be meas..ed fran <letector. output or input of first. electronic conponent'in channel.

i

TABLE 4.3-1 REACTOR PROTECTI0fl SYSTEM INSTRUMENTATE0N SURVEILLANCE REQUIREllENTS n

C}{ANNEL -

MODES IN WHIC}l y

CilAtlNEL CHANNEL

' FUNCTI0flAL SURVEILLANCE FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED h

1.

Manual Reactor Trip FL A.

fl. A.

S/U(1)

N.A.

2.

Nuclear Overpower S

D(2) and Q(7).

M 1, 2 3.

RCS Outlet Temperature--Iligh

-S R

H 1, 2

[

4.

Fluclear Overpower Based on RCS Flow and AXIAL POWER. IMBALANCE S(4)

M(3) and Q(7,8) M 1, 2 5.

RCS Pressure--Low S

R H

1, 2 6.

RCS Pressure--High S

R H

1,.2 7.

Variable Low RCS Pressure S

R' M

1, 2 8.

Reactor Containment Pressure--liigh S R

H 1, 2 m

S 9.

Intermediate Pange, Neutron Flux and Rate S

R(7)

S/U(1)(5) 1, 2 and *

10. Source Raage, Neutron Flux and Rate S

R(7)

S/U(1)(5) 2, 3, 4 and 5

11. Control Rod Drive Trip Breaker N.A.

N.A.

M and S/U(1) 1, 2 and *

[

12. Reactor Trip Module N.A.

N.A.

M 1, 2, and

• h 13.

Shutdown Bypass RCS S

R M

2**,

3**, 4**, 5**

g Pressure-liigh l;

14. Reactor Coolant Punp Power Monitors S R

H 1, 2 e

1 4

m-s e. De I

f.

-I TABLE 4.3-1 (Continued) f0TATION With any control rod drive trip breaker closed.

When Shutdown Bj pss is actuated.

(1)

If not performed in previous 7 days.

(2)

Heat balance only, above 15% of RATED THERMAL POWER.

(3)

Wnen THERMAL POWER [TP) is above 30% of RATED THEPf4AL POWER [RTP), compare out-of-core measured AXIAL PCWER IMBALANCE [ API ] to incore measured AXIAL POWER IMBALANCE

[ API ]. as follhs:

y RTP (API API ) = Imbalance Error U

g TP Recalibrate if the absolute value of the Imbalance Error is equal to or greater than 3,5%.

(4) -

AXIAL POWER IMBALANCE and 1000 flow indications only.

Verify at least one decade overlap if not verified in previous (5) 7 days.

Each train tested every other month.

(6)

(7) -

I;eutron Jetectors may be excluded from CHANNEL CALIBRATION.

Flow rate measurement sensors may be excluded from CHANNEL (8)

CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 months.

CRiSTAL RIVER - UNIT 3 3/4 3-8 Amendment No. 41

,M "o

UNITED STATES

+

P

~, ',

NUCLEAR REGULATORY COMMISSION

{

,I ADVlsCRY COMMITTEE ON REACTOR SAFEGUARDS 8,

g WASHINGTON, c. C. 20558

%,

• v

,e, i

May 13, 1981 MEMCRANDUM FOR: William J. Of rcks, Executive Director for Operations FROM:

Raymond F. Fraley x ut ve Director, ACRS

SUBJECT:

POWER LEVEL INCREASE AT CRYSTAL RIVER NUCLEAR PLANT UNIT 3 During its 253rd meeting, the ACRS heard a report from its Subcommittee on Babcock and Wilcox reactors regarding the request frcm the Florida Power Corporation to increase the power level of the Crysta! River Nuclear Plant from 2452 MW: to 2544 MWt. This request was discussed by the Subcommittee during a meeting in Washington, DC on May 6,1981.

Based on this report the Committee concluded that it need not review fur-ther the proposed power level increase and has no objection to the NRC Staff licensing the Crystal River Nuclear Plaiit Unit 3 te operate at power levels up to 2544 MWt.

e h

~

' g.

p 4

%:., % d W

ye.

A D t) P o F 8105210,233 e

,