ML20003G009

Amend 38 to License DPR-72,revising Tech Specs to Be Consistent W/Lessons Learned Category a Requirements in Response to Revised

ML20003G009
Person / Time
Site:	Crystal River
Issue date:	04/17/1981
From:	Stolz J Office of Nuclear Reactor Regulation
To:	City of Alachua, FL, City of Bushnell, FL, City of Gainesville, FL, City of Kissimmee, FL, City of Leesburg, FL, City of New Smyrna Beach, FL, City of New Smyrna Beach, FL, Utilities Commission, City of Ocala, FL, City of Orlando, FL, City of Tallahassee, FL, Florida Power Corp, Orlando Utilities Commission, Sebring Utilities Commission, Seminole Electric Cooperative
Shared Package
ML20003G010	List: ML20003G009 ML20003G013 ML20003G017
References
DPR-72-A-038 NUDOCS 8104280052
Download: ML20003G009 (46)

Text

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

,E wAsHWGTON. D. C. 20686

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FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNEi.L CITY 0F GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF NEW SHYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO.

SEBRING UTILITIES COMMISSION SEMINOLE ELECTRIC COOPERATIVE, INC.

~

CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT

' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 38 License No. DPR-72 1.

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

A.

The application for amendment by Florida Power Corporation, et al. (the licensees) dated September 15, 1980, as revised December 31, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set f';rth in 10 CFR Chapter I; B.

The facility will operate in conformity J.th the application, the provisions, of the Act, and the rule 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 Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E..The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

J P

8104280052

, i 2.

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

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.38 are hereby incorporated in the license.

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

3.

This license amendment is effective as of thc date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION s

Jo W F. -Stolz, Chief gj 0 rating Reactors Branch #4 vision of Licensing e

attachment:

Changes to the Technical Specifications Jate of Issuance: April 17,1981 t

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ATTACHMENT TO LICENSE AMENDMENT NO. an FACILITY OPERATING LICENSE NO. DPR 72' DOCKET NO. 50-302 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to raintain document completeness.

Paoes IV 3/4 4-3 3/4 3-10 3/4 4-4 3/4 3-11 3/4 4-5 3/4 3-12 3/4 6-17 3/4 3-13 3/4 6-18 3/4 3-14 3/4 6-19

'3/4 3-15 3/4 6-20 f

3/4 3-16 3/4 6-21 3/4 3-16a (new page) 3/4 6-21a (new page) 3/4 3-17 3/4 7-5 3/4 3-17a (new page)

B 3/4 4-1 l

3/4 3-18 B 3/4 4-2 l

3/4 3-19 B 3/4 4-3 l

3/4 3-20 6-4 3/4 3-38 6-5 3/4 3-39 6-12

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REnUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL S h utdown Ma rgi n - O pe ra ti n g..........................

3/4 1-1 Shutdown Ma rgi n.. S hutdown...........................

3/4 1-2a B a ro n D i l u t i o n.......................................

3/4 1-3 i

Moderator Temoerature Coefficient....................

3/4 1-4 Minimum Temocrature for Cri ticality.................. 3/4 1-5 3/4.1.2 BORATION SYSTEMS t

F l ow P a th s - S h ut d own................................ 3/4 1-6 iU Fl ow Pa th s - Ope rati n g...............................

3/4 1-7 I~

Ma k eu p Pumo - S hutdown............................... 3/4 1-9 Makeup Pumps -

0oerating.............................

3/4 1-10 Decay He a t Removal Pumo - Shutdown...................

3/4 1-11 Boric Acid Pumo - Shutdown...........................

3/4 1-12 Bori c Aci d P umos - Ope ra ti n g.........................

3/41'.3 Sorated Water Sources - Shutdown.....................

3/4 1-14 Borated Water Sources - 0oerating....................

3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height - Safety and Regulating Rod Groues......

3/4 1-18 Group Height - Axial Power Shaoing Rod Grouo.........

3/4 1-20 Position Indi cator Channel s..........................

3/4 1-21 R o d D ro o T i me........................................

3/4 1-23 Safety Rod Insertion Limit...........................

3/4 1-24 l

Regulating Rod Insertion Limits......................

3/4 1-2S Rod Program..........................................

3/4 1-33 l

' Axial Power Shapina Rod Insertion Limits..............

3/4 1-37 i

CRYSTAL RIVER - UNIT 3 III Amendment No. 16, 32, 34 s

l

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEIL 1.ANCE ' REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXI AL POWER I MBALANCE............................... 3/4 2-1 3/4. 2. 2 NUCLEAR HEAT FLUX H0i' CHANNEL FACTOR - Fg........... 3/4 i-4 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FN...... 3/4 2-6 AH 3/4.2.4 QUADR ANT POWER TI LT................................. 3/4 2-8 3/4.2.5 DNB PARAMETERS...................................... 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM I NSTRUME NTATI ON........... 3/4 3-1 3/4. 3.~2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM

, I NSTR UME NTATI ON................................... 3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION Radi ation Moni to ri ng Instrumentation................ 3/4 3-22 In core Detecto rs.................................... 3/4 3-26 Sei smic Instrumentation............................. 3/4 3-28 Meteorol ogical Instrumentation...................... 3/4 3-31 Remote Shutd own Instrumentation......'............... 3/4 3-34 Post-acci de nt Instrumentation....................... 3/4 3-37 Fi re Detectio n Instrument ation...................... 3/4 3-40 c

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 R EACTOR COOLANT L00PS..................'............. 3/4 4-1 3/4.4.2 RELIEF VALVES - SHUTD0WN............................ 3/4 4-3 3/4.4.3 RELIEF VALVES - 0PERATING........................... 3/4 4-4 Code Sa fe ty Val ves.................................. 3/4 4-4 Power-Ope rated Rel i ef Val ve......................... 3/4 4-4a CRYSTAL. RIVER - UNIT 3 IV Amendment No. JJ, 38

-.. = - -

INSTRUMENTATION g

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERA _ TION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumen-

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tation channels shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an ESFAS instrumentation channel trip setpoint less a.

conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint Value.

b.

With an ESFAS instrumentation channel inoperable, take the action shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS Each ESFAS instrumentation channel shall be demonstrated 4.3.2.1.1 OPERABLE by the performance of the CHANNEL CHECR, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations durin; the MODES and at the frequencies shown in Table 4.3-2.

The logic for the bypasses shall be demonstrated OPERABLE 4.3.2.1.2 during the at power CHANi,-L FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFA 4.3.2.1.3 function shall be demonstrated to be within the limit at least once per Each test shall include at least one channel per function 18 months.

such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

CRYSTAL RIVER - UNIT 3 3/4 3-9 e,,.

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TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION n"-<

4 4#

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE m

2 FUNCTIONA:. UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 9

1.

SAFETY INJECTION a.

High Pressure Injection 1.

Manual Initiation 2

1 2

1,2,3,4 13 2.

Reector Bldg. Pressure Hi@

3 2

2 1,2,3 9#

3.

RCS Pressure Low 3

2 2

1, 2, 3*

9#

4.

RCS Pressure Low-Low 3

2 2

1, 2, 3**

9#

5.

Automatic Actuation Logic 2 1

2 1,2,3,4 10 b.

Low Pressure Injection w

1.

Manual Initiation 2

1 2

1,2,3,4 13 l

2.

Reactor Bldg. Pressure i

High 3

2 2

1,2,3 9#

3.

RCS Pressure Low-Low 3

2 2

1, 2, 3**

9#

4.

Automatic Actuation Logic 2 1

2 1,2,3,4 10 2.

REACTOR BLDG. C00 LIT, a.

Manual Initiation 2

1 2

1,2,3,4 13 k

b.

Reactor Bldg. Pressure a

High 3

2 2

1,2,3 9#

z c.

Automatic Actuation Logic 2

1 2

1,2,3,4 10 j

8

3 TABLE 3.3-3 (Cont.'d)

W ENGINEERED SAFETY FEATURE ACTUATION, SYSTEM INSTRUMENTATION _

r l

E'.

i M

MINIMUM y

TOTAL NO.

CllANNELS CilANNELS APPLICABLE E

FUNCTIONAL UNIT _

0F CilANNELS TO TRIP TERABLE_

MODES ACTION _

a.

4 3.

REACTOR BLOG. SPRAY i

to a.

Reactor Pldg. Pressure i

liigh-liigh coincident 4

with liPI Signal 3

2 2

1,2,3 12 4

b.

Automatic Actuation Logic 2

1 2

1,2,3 10 to h

4.

0 tiler SAFETY SYSTEMS Reactor Bldg. Purge E haust Duct '

x 1

a.

Isolation on Illgh Radioactivity i

1 1

1 1,2,3,4 lif Gascous 4

!!?

N i

a I

s lI D

l j

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TABLE 3.3-3(Cont'd)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INS _TRUMENTATION t

g)

MINIMUM

• II TOTAL NO.'

CHANNELS CllANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE

_ MODES ACTION 5

=

1 j{

b.

Steam Line Rupture Matrix w

i 1.

Low SG' Pressure 2 per steam 1 per steam 2 per steam 1, 2, 3***

10 generator generator generator gj 2.

Automatic Actuation Logic 1 per steam l'per steam 1 per steam I, 2, 3 Id j

1 generator generator generator 4

I c.

Emergency Feedwater 1.

Main Feedwater Pump Turbines 2

1, 2, 377 10 A and B Control Oil Low 2

2 u,

2.

OTSG A and B Level Low-Low 2

2 2

1, 2, 3, 4 10 u,

b a

R a

M

TABLE 3.3-3 (Cont'd)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEf INSTRUMENTATION c

r-MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE x

9 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION h

5.

REACTOR BLDG. ISOLATION a.

Manual Initiation 2

1 2

i l, 2, 3, 4 13 b.

Reactor Bldg. Pressure.

High 3

2 2

1,2,3 9#

c.

Automatic Actuation Logic 2

1 2

1,2,3,4 10 d.

Manual Initiation 2

1 2

1,2,3,4 13 (HPI Isolationi) w e.

RCS Pressure Low 3

2 2

1, 2, 3*

13 (HPI Isolation) f.

Automatic Actuation Logic 2

1 2

1,2,3,4 10 (HPI Isolation) l 8

F a

i

l TABLE 3.3-3 (Continued)

TABLE NOTATION

• Trip function may be bypassed in this MODE with RCS pressure below 1700 psig.

Bypass shall be automatically removed when RCS pressure exceeds 1700 psig.

• • Trip function may be bypassed in this MODE with RCS pressure below 900 psig.

Bypass shall be automatically remo~ved when RCS pressure exceeds 900 psig.*

• • • Trip function may be bypassed in this MODE with steam generator pres-sure below 725 psig.

Bypass shall be automatically removed when steam generator pressure exceeds 765 psig.

1. The provisions of Specification 3.0.4 are not applicable.

1. 1. Trip function may be bypassed in this MODE prior to stopping the operating main feedwater pump.

Bypass shall be manually removed after starting the first main feedwater pump.

ACTION STATEMENTS ACTION 9 -

With the nunber of OPERABLE Channels one less than the Total Number of Channels operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the l

inoperable channel is placed in the tripped condition within i

1. hour.

ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, 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 /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to I hour for surveillance testing per Specification 4.3.2.1.1.

ACTION 11 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves are maintained closed, l

ACTION 12 - With the number of OPERABLE Channels one less than the Total Number of Channels operation may proceed provided the inoper-able channel is placed in the bypassed condition and the minimum channels OPERABLE required is demonstrated within I hour; 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 4.3.2.1.

ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of

Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STAN0BY 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 fol-l lowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

CRYSTAL RIVER - UNIT 3 3/4 3-14 Amendment No. 38

i l

i.

TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETPOINTS i

5

o FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES G

9 1.

SAFETY INJECTION a.

High Pressure Injection ES-Actuation "A" and "B" c-z 1.

Manual Initiation Not Applicable Not Applicable 2.

Reactor Bldg. Pressure High 1 4 psig

'f 4 psig 1500 psig 3.

RCS Pressure Low-

> 1500 psig

),500 psig t

4.

RCS Pressure Low-Low T 500 psig

)

5.

Automatic Actuation Logic Not Applicable Not App 1tcable w

b.

Low Pressure Injection ES 1

Actuation "A" and "B" wL 1.

Manual Initiation Not Applicable Not Applicable 2.

Reactor Bldg. Pressure High I 4 psig f 4 psig 3.

RCS Pressure low-Low

> 500 psig

) 500 psig 4.

Automatic Actuation Logic Not Applicable Wot Applicable 2.

REACT (R BLDG. COOLING a.

ES Actuation "A" and "B" E

1.

Manual Initiation Not Applicable Not Applicable 2.

Reactor Bldg Pressure 111gh

< 4 psig

< 4 psig i

3.

Automatic Ac aation Logic Wot Applicable Not Applicable j

5 b.

ES Actuation Indication "AB"

)

M 1.

Automatic Actuation Logic Not Applicable Not Applicable i

4

TABLEc3.3-4 (Cont'd)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION TRIP SETPOINTS g

FUNCTIONAL UNIT TRIP.SETPOINT:

ALLOWABLE VALUES

= 3.

REACTOR BLDG. SPRAY

)

E a.

Reactor Bldg. Pressure.

High-High

< 30 psig

< 30 psig l

e coincident with HPI Signal Tee 1.a.2, 3, 4 Tee 1.a.2, 3, 4 cx I

b.

Automatic Actuation Logic Not Applicable Not Applicable l

4.

OTHER SAFETY SYSTEMS a.

Reactor Bldg. Purge Exhaust Duct

,i Isolation on High Radioactivity Not Applicable Gaseous b.

Steam Line Rupture Matrix i

1.

Low SG Pressure 1 600 psig 1 600 psig 4

2.

Automatic Actuation Logic Not Applicable Not Appifcable e

c.

Emergency feedwater

[

7j 1.

Main Feedwater Pump Turbines > 55 psig 1 55 psig i

g A and B Control Oil Low i

g 2.

OTSG A and B Level Low-Low 1 18 inches 1 18 inches 4

s

$

• Determined by requirements of Appendix "B" Tech. Specs. Section 2.4.2 - Crystal River 3 Operating License No. OPR-72.

i l

i

~

i TABLE 3.3-4 (Cont'd) n E

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS A

3 FUNCTIONAL UNIT

~

TRIP SETPOINT ALLOWABLE VALUES g

f 5.

REACTOR BLDG. ISOLATION I

E a.

ES Actuation "A" and "B" ia 1.

Manual Initiation Not Applicable Not Applicable t

2.

Reactor Bldg. Pressure High

< 4 psig

' < 4 psig i

3.

Automatic Actuation Logic Wot Applicable Wot Applicable j

4.

Manua s Inttiation 1

(IIPI Isolation)

Not Appilcable Not Applicable

~

to 5.

RCS Pressure Low i

1 (HPI isolation)

< 1500 psig

< 1500 psig to 6.

Automatic Actuation Logic l

g (HPI Isolation)

Not Appifcable Not Appifcable l

4 i

I k

l i

n l

i l

1 i.

4 e

j 4

4

O TABLE 3.3-5 ENGINEERED _, SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

• 1.

Manual a.

High Pressure Injection Not Applicable b.

Low Pressure Injection Not Applicable c.

Reactor Building Cooling Not Applicable d.

Reactor Building Isolation Not Applicable e.

Reactor Building Spray Not Applicable f.

Reactor Building Purge Isolation Not Applicable 9

Steam Line Rupture Matrix 1.

Emergency Feedwater Actuation Mot Applicable 2.

Feedwater Isolation Not Applicable 3.

Steam Line Isolation Not Applicable h.

HPI Isolation Not Applicable 2.

Reactor Building Pressure-High a.

High Pressure Injection 25*

b.

Low Pressure Injection 25*

c.

Reactor Building Cooling 25*

d.

Reactor Building Isolation 60*

3.

Reactor Building Pressure High-High (with HPI signal)

. a.

Reactor Building Spray 56*

4.

RCS Pressure Low a.

High Pressure Injection 25*

b.

HPI Isolation 60*

5.

RCS Pressure Low-Low a.

High Pressure Injection 25*

b.

Low Pressure Injection 25*

6.

Low Steam Generator Pressure a.

Feedwater Isolation 34 b.

Steam Line Isolation 5

i CRYSTAL. RIVER - UNIT,3 3/4 3-17 Amendment No. 38

TABLE 3.3-5 (Cont'd)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

• 7.

Containment Radioactivity-High a.

Reactor Building Purge Isolation 15

• 8 Main Feedwater Pump Turbines A and B Control Oil Low a.

Emergency Feedwater Actuation Not Applicable 9

OTSG A and B Level Low-Low a.

Emergency Feedwater Actuation Not Applicable T

• Diesel Generator starting and sequence loading delays included.

Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

CRYSTAL RIVER - UNIT 3 3/4 3-17a Amendment No. 38

TABLE 4.3-2 n

i ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

"<4 i

i A

CllANNEL MODES IN WHICH

o CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE y

o FUNCTIONAL UNIT CilECK CAllBRATION TEST REQUIRED l

l g

1.

SAFETY INJECTION a.

High Pressure Injection 1.

Manual Initiation N/A N/A M(1) 1, 2, 3, 4 2.

Reactor Bldg. Pressure High S

R M(2) 1, 2, 3 u

i 3.

RCS Pressure Low S

R H

1, 2, 3 i

4.

RCS Pressure Low-Low S

R M

1, 2, 3 5.

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3, 4 I

i b.

Low Pressure injection f

1.

Manual Initiation N/A N/A M(1) 1, 2, 3, 4 2.

Reactor Bldg. Pressure High S

R M(2) 1, 2, 3 j

i 3.

RCS Pressure low-low S

R M

1, 2, 3 4.

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3, 4 3

k i

2.

REACTOR BLDG. COOLING a

a i

I a.

Manual Initiation N/A N/A M(1) 1, 2, 3, 4 h

b.

Reactor Bldg. Pressure s

R h(2')

1, 2, 3 High c.,

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3, 4

^..

TABLE 4.3-2 (Cont'd)

ENGINEER _ED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SU P

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SilRVEILLANCE q

CHECK CALIBRATION TEST RE_ QUIRED q FUNCTIONAL UNIT g 3.

REACTOR BLDG. SPRAY

[

a.

Reactor Bldg. Pressure High-High coincident with HPI Signal S

R M(4) 1, 2, 3 b.

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3 t'

[4.

OTIIER SAFETY SYSTEMS a.

Reactor Bldg. Purge Exhaust Duct Isolation on High Radioactivity 1.

Gaseous s

Q M

All Modes b.

Steam Line Rupture Matrix 1.

Low SG Pressure N/A R

N/A 1, 2, 3 2.

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3 k

c.

Emergency Feedwater i

5 l

1.

Main Feedwater Pump Turbines S R

N/A 1,2,3 2

P A and B Control Oil Low 2.

OTSG A and B Level Low-Low S

R N/A 1,2,3,4 e

m- -

TABLE 4.3-2 (Cont'd) i n

l

}

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS h

CHANNEL MODES IN WHICH i

o j

Q CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED i

1 5.

REACT (R BLDG. ISOLATION j

g

~

a.

Manual Initiation N/A N/A M(1) 1, 2, 3, 4

]

b.

Reactor Bldg. Pressure High S

R M(2) 1, 2, 3 c.

Automatic Actuation Logic N/A N/A M(3) 1, 2, 3, 4 d.

Manual Initiation (HPI Isolation)

N/A N/A M(1) 1, 2, 3, 4 j

e.

RCS Pressure Low (HPI Isolation)

S R

M 1, 2, 3 f.

Automatic Actuation Logic (HPI Isolation)

N/A N/A M(3) 1, 2, 3, 4 O

a a

a D

i l

\\

INSTRUMENTATION POST-ACCIDENT INSTRUMENTATION LIM

• TING CONDITION FOR OPERATION 3.3.3. 6 The post-accident monitoring instrunentation channels shown in Tatie 3.3-10 shall be OPERAS ~ E with readouts and reccrders in the control room.

AD3LICAE:t:TY: MODES 1, 2 an:: 3.

ACTION:

With the number of OCERAELE post-acticer.: monitoring channels a.

less *han re:;uired by Tatie 3.3-10, either restore tne inoperatie channel to OPER*E'.- s:atus within 30 days, or be in HOT SHUTSi within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

The provisions of Specification 3.0.1 are n:: applicable.

SURVEILLANCE REQUIREME'iTS Each post-acciden; monitoring instrumentation channel shall be 4.3.3.5 demonstrated OPERABLE by perfomance of tne CHANNEL CHECK and CHANNEL CALISRATION operations at the frecuencies shown in Table 4.3-7.

i l

l l

l l

l l

l l

[

3/4 3-37 i

CRYSTAL RIVER - UNIT 3 e

-.,,m,,-r

,c.

e<-

.J i

I TABLE 3.3-i0 n

POST-ACCIDENT MONITORING INSTRUMENTATION P

MINIMUM m

MEASUREMENT CHANNELS

]9 RANGE OPERABLE INSTRUMENT 0-125%

2

[

1.

Power Range Nuclear Flux 2.

Reactor Building Pressure 0-70 psia 2.

10-1 to 106 cps 2

w 3.

Source Range Nuclear Flux 4.

Reactor Coolant Outlet Temperature 520 *F

.620 *F 2 per loop 5.

Reactor Coolant Total Flow 0-110% full flow 1

ginb?2SbO'psig

{

g500puig 6.

RC Loop Pressure l

2 M.

0-320 inches 2

y 7.

Pressurizer Level m

8.

Steam Generator Outlet Pressure 0-1200 psig 2/ steam generator 9.

Steam Generator Operating Range Level 0-1005 2/ steam generator

10. Borated Water Storage Tank Level 0-50 feet 2

0-1.5x106 lb/hr.

2 11.

Startup Feedwater Flow

'12.

Reactor Coolant System Subcooling Margin Monitor

-658 F* to +668 F*

1 g

13.

PORV Position Indicator (Primary Detector)

N/A 1

k 14.

PORV Position Indicator (Backup Detector)

N/A 0

15. PORV Block Valve Position Indicator N/A 0

16.

Safety Valve Position Indicator (Primary Detector)

N/A 1/ Valve o

I e Position Indicator (Backup 17.

Sagy 1

Y q

~

~

i TAfsLE 4.3-7 9

POST-ACCIDENT MONITORING INSTRU ENTATION SURVEILLANCE REQUIREMENTS 1

)i U

hf' P

~~

CHANNEL CHANNEL CO CHECK CALIBRATION m

INSTRtMENT j

M Q*

1. CPower Range Nuclear Flux M

R l

2.., Reactor Building Pressure

.... )

M R*

3.;.,S.ource Range Nuclear Flux H

M R

w 4.+. LReactor Coolant Outlet Temperature M

R l

5. AReactor Coolant Total Flow Rate M

R 6.

RC Loop Pressure M

R 7.

Pressurizer Level w

M R

l 8.

Steam Generator Outlet Pressure w

M R

]

g 9.

Steam Generator Level M

R 10.

Borated Water Storage Tank Level i

M R

11.

Startup Feedwater Flow Rate

~

M R

12. Reactor Coolant System Subcooling Margin Monitor M

R 13.

PORY Position Indicator (Primary Detector) 14.

PORY Position Indicator (Backup Detector)

M R

M R

{

15.

PORY Block Valve Position Indicator y 16.

Safety Valve Position Indicator (Primary Detector)

M R

l 17.

Safety Valve Position Indicator (Backup Detector)

M R

I k

• Neutron detectors may be excluded from CHANNEL CALIBRATION.

1 1

4 4

INSTRUMENTATION FIRE DETECTION INSTRUM.ENTATION LIMITING CONDITION FOR OPERAT_1_0 3.3.3.7 As a minimum, the fire' detection instrumentation for each fire detection zone shown ir. Table 3.3-11 shall be OPERABLE.

APPLICABILITY:

Whenever equipment in that fire detection zone is required to oe' OPERABLE.

ACTION:

With one or more of the fire detection instrument (s) shown in Table 3.3-11, inoperable:

a.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish a fire watch patrol to inspect the zone (s) witt. the inoperable instrur.ent(s) at least once per hour, and b.

Restore the inoperable instrurent(s) tc 0:EKABLE status within 14 days or ir. lies of any other report required by Specification-6.9.1, prepare and subr.it a Speciel Report to the Commission pursuant to Specification 6.9.2 within the next 30 days out-lining the action taken, the cause of the malfunction and the plans and schedule for restoring the instrument (s) to OPERABLE status.

c.

The ' provisions of Specifications 3.0.3 and 3.0.4 are not aoplicable.

SU.VEILLANCE REOUIREMENTS 4.3.3.7.1 Each of the above fire detection instruments shall be demonstrated 0 ERAELE at least on e per 6 months by performance of a CHANNEL FUNCTIONAL TEST.

4.3.3.7.2 The circuitry tssociated with the detector alarms listed in Tatle 3.3-11 shall be demonstrated 0 ERAELE at least once per 6 months for all National Fire Frctection Association (NFFA) Coos 72D Class 5 supervised circuits.

4.3.3.7.3 The non-supervised circuits between the local panels and the control room for the detectors listed in Table 3.3-11 shall be demonstrated 00ERABLE at least once per 31 days.

P00R ORIGIN 4 CRYSTAL RIVER - UNIT 3 3/4 3-40 Amendment No. II 8

REACTOR COOLANT SYSTEM RELIEF VALVES - SHUTDOWN CODE SAFETY VALVES l

LIMITING C021 TION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 psig +,1%.

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE DHR loop into operation.

1 SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

Amendment No. 38 CRYSTAL RIVER - UNIT 3 3/4 4-3

REACTOR COOLANT SYSTEM RELIEFVALVES-OPERATIQ CODE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.3.

All pressurirar code safety valves shall be OPERABLE with a lift setting of 2500 psig + 1%.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.3 No additional Iurveillance Requirements other than those required by Specification 4.0.5.

Amendment No. 38 CRYSTAL RIVER - UNIT 3 3/4 4-4

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR.0PERATION 3.4.4 The pressurizer shall be OPERABLE with:

a.

A steam bubble, b.

A water level between 40 and 290 inches, and c.

At least 126 kW of pressurizer heaters.

APPLICABILITY: MODES I and 2.

ACTION:

With the pressurizer inoperable, be in at least HOT STANDBY with the control rod drive trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.4.1 The pressurizer shall be demonstrated OPERABLE by verifying pressurizer level to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by manually transfer-ring power from the normal to the emergency power supply and energizing the heaters.

i j

CRYSTAL RIVER - UNIT 3 3/4 4-5 Amendment No. 38 m

REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITING CONDITION F0'R OPERATION 3.4.5 Each steam generator shall be OPERABLE with a water level between 18 and 360 inches.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With one or more steam generators inoperable due to steam generator tube imperfections, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200*F.

b.

With one or more steam generators inoperable due to the water level being outside the limits, 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 /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by oerformance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4. 4. 5.1.

Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

i The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4 4.5.4.

The tubes selected for each inservice inspection shall include at least 3%

of the. total numbar of tubes in all steam generators; the tubes selected I

for these inspections shall be selected on a random basis except:

Where experience in similar plants with similar water chemistry a.

indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

CRYSTAL' RIVER - UNIT 3 3/4 4-6

TABLE 3.6-1 CONTAINMENT IS9tATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds)

A.

CONTAINMENT ISOLATION 1.

B5V-27 check #

closed dur. nor. operation NA l

and open dur. RB spray 60 BSV-3 #

NA-BSV-26 check #

60 BSV-4 #

2.

CAV-126(A) iso. CA sys. fr. RC letdn.

60 CAV-1**(A) iso. CA sys. fr. pzr.

60 60 CAV-3 (A)

CAV-2 (B) iso. CA sys. fr. RB 60 CAV-4 # (A) isolate liquid sampling 60 system 60 CAV-6 i (B) 60 CAV-5 # (A) 60 CAV-7 # (B) 3.

CFV-20 check iso. N2 supply fr. CFT-1A NA 60 CFV-28 (A/B)

NA CFV-17 check iso. N2 supply fr. CFT-1B 60 CFV-27 (A/B)

CFV-18 check iso. MU system fr. CFT-1B NA 60 CFV-26 (A/B)

CFY-19 check iso. MU system fr. CFT-1A NA 60 CFV-25 (A/B)

CFY-42 (B) iso. liquid sampling fr.

60 CF system CFV-15 (A) iso. WD sys. fr. CF tanks 60 60 CFV-16 (A) 60 CFV-29 (B)

CFV-11 (A) iso. CF tanks fr. liquid 60 sampling system 60 CFV-12 (A)

CRYSTAL RIVER - UNIT 3 3/4 6-17 Amendment No. 27,38

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES VALVE NUMBEk FUNCTION ISCLATION TIME (seconds) 4.

CIV-41 iso. CI sys. fr. RB 60 60 CIV-40 60 CIV-34 60 CIV-35 5.

DHV-93 check iso. DH system fr. pzr.

NA 60 CHV-91 DHV-43 #

iso. DH sys. fr. RB sump 120 DHV-42 #

120 DHV-4f & 41#

iso. DH sys. fr. RC 120 DHV-6 #

iso. DH system from 60 Reactor Vessel 60 DHV-5 #

6.

DWV-162 check iso. DW system fr. RB NA 60 DWV-160 (A/Si 7.

FWV-44 check i iso. feedwater NA from RCSG-1A NA FWV-45 check #

FWV-43 check #

iso. feedwater NA from RCSG-1B NA FWV-45 check #

8.

MSV-130 #(A/B).

iso. MDT-1 fror.RCSG-1A 60 MSV-148 f(A/B) iso. MDT-1 from RCSG-1B 60 l

MSV-411 i iso. main steam lines 60 l

from RCSG-1 A MSV-412 #

iso. main steam lines 60 from RCSG-1A MSV-413 #

iso. main steam lines 60 from RCSG-1B MSV-414 #

iso main steam lines 60 from RCSG-18 I

Amendment No.38 CRYSTAL RIVER - UNIT 3 3/4 6-18 i

~ -,

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds) 9.

MUV-40 (A) iso. MU system from RC 60 MUV-41 (A) 60 MUV-49 (B)-

60 MUV-253 60 MOV-261 iso. MU system from 60 control bleed-off MUV-260 60 MUV-259 60 MUV-258 60 MUV-163 check #

open durinj HPI and NA closed dur. nor. operation l

MUV-25 i 60 MUV-164 check #

NA MUV-26 #

60 MUV-160 check #

NA MUV-23 #

60 MUV-161 check #

NA MUV-24 #

60 MUV-27 #

open dur. nor. operation 60 and closed during RB Isolation

10. SWV-39 #

iso. NSCCC from AHF-1C 60 SWV-45 #

60 SWV-35 #

iso. NSCCC from AHF-1A 60 60 SWV-41 #

SWV-37 #

iso. NSCCC from AHF-1B 60 SWV-43 #

60 SWV-48 i isolate NSCCC fraa 60 MUHE-1A & IB and WDT-5 60 SWV-47 #

60 SWV-49 #

60 SWV-50 #

SWV-80 #

iso. NSCCC from RCP-IA 60 60 SWV-84 #

SWV-82 #

iso. NSCCC from RCP-IC 60 60 SWY-06 #

CRYSTAL RIVER - UNIT 3 3/4 6-19 Amendment No. 77, 24,38

~

4 TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds)

SWV-81 #

iso. NSCCC from RCP-1D 60 60 SWV-85 #

'SWV-79 #

iso. NSCCC from RCP-1B 60 60 SWV-83 #

SWV-109#

iso. NSCCC from DRRD-1 60 60 SWV-110#

11. WDV-4 (B) iso. WDT-4 from RB sump 60 60 WDV-3 (A)

WDV-60 (A) iso. WDT-4 from WDT-5 60 60 WDV-61 (B)

WDV-94 (A) iso. WDT-4 from WDP-8 60 60 l

WDV-62 (B )~

WDV-406 (A) iso. waste gas disposal 60 from vents in RC system 60 WDV-405 (B) 12.

WSV-3 iso. containment monitoring 60 system from RB 60 WSV-4 60 WSV-5 60 WSV-6 B.

CONTAINMENT' PURGE AND EXHAUST 1.

AHV-1C (A iso. pur. sup. system fr. RB 60 60 AHV-10 (B

/.HV-1B (A) iso. pur. exhaust system fr. RB 60 60 AHV-1A (B)

C.

MANUAL 1.

IAV-28 iso. IA from RB NA NA IAV-29 2.

LRV-50 iso. leak rate test system NA from RB NA LRV-36 CRYSTAL RIVER - UNIT 3 3/4 6-20 Amendment No. 38

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds)

LRV-51 iso. atmos. vent and RB NA

~

LRV-35 & 47 purge exhaust system from RB NA LRV-49 iso. atmos. vent from RB NA LRV-38 & 52 NA LRV-45 iso. LR test panel from RB NA LRV-44 NA LRV-46 NA 3.

MSV-146#

iso. misc. waste storage NA tank from RCSG-1B 4.

NGV-62 iso. NG system from NA steam generators NA NGV-81 #

NGV-82 isc. NG system from pzr.

NA 5.

SAV-24 iso. SA from RB NA NA SAV-23 & 122 6.

SFV-18 iso. SF system NA NA SFV-19 SFV-119#

iso. Fuel Transfer tubes NA from F.T. Canal NA SFV-120f 4

7.

WSV-1 iso. containment monitoring NA system from RB NA WSV-2 D.

PENETRATIONS REQUIRING TYPE B TESTS l

Blind Flange 119 iso. RB NA NA Blind Flange 120 NA Blind Flange 202 i

4 f

~

l CRYSTAL RIVER - UNIT 3 3/4 6-21 Amendrent No. 27, 28, 38

TABLE 3.6-1 (Continued)

CONTAltMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME (seconds)

Blind Flange 348 iso. fuel transfer tube from NA Transfer Canal NA Blind Flange'436 Equipment Hatch iso. RB NA Personnel Hatch iso. RB NA i Not subject to Type C Leakage Test

• • The provisions of Specificat on 3.0.4 are not applicable until i

startup for Cycle 3 operation.

Isolation valves closed to of Specification 3.6.3.1 ACTION b. and satisfy the requirements

c. may be re-opened on an intermittent basis under administrative control for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period as necessary for sampling.

l (A)

Isolates on Diverse Isolation Actuation Signal A Isolates on Diverse Isolation Actuation Signal B l

(B)

(A/B) Isolates on Diverse Isolation Actuation Signal A or-B e

I e

l CRYSTAL RIVER - UNIT 3 3/4 6-21a Amendment No.38

CONTI.INMENT SYSTEMS 3 /4. r;. 4 COMBUSTIBLE GAS CONTROL MYDROGEN ANALYZERS LIMITING CONOITION FOR OPERATION 3.6.4.1 A containmen't hydrogen analyzer and a gas chromatograph shall be OPERABLE.

A?olICAEILITY: MODES 1 and 2.

ACTION:

With one hydrogen analysis device inoperable, restore the inoperable device to OPERABLE status within 30 days or be in at least HDT STAND 3Y within the nex: 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REOUIREMENTS l

4.6.4.1 Ea:h hydrogen analysis device shall be demonstrated OPERABLE l

at least once per 92 days on a STAG 3ERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:

l One volume per:ent hydrogen, balance nitrogen, and a.

Four volume percent hydrogen, balan:e nitrogen.

b.

CRYSTAL RIVER - UNIT 3 3/4 6-22

PLANT SYSTEMS i

SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

3.

Verifying that the emergency feedwater ultrasonic flow rate detector is zero-chet.ked.

b.

At least once per 18 months, during shutdown,' by:

1.

Verifying that each automatic valve in the flow path actuates to its ccerect position on an eme gency feed-water actuation test signal.

2.

Verifying that the steam turbine driven pump and the motor driven pump start automatically:

Upon receipt of an emergency feedwater actuation l

a.

OTSG A and B level low-low test signal, and b.

Upon receipt of an emergency feedwater actuation raain feedwater pump turbines A and B control oil low test signal.

3.

Verifying that the operating air accumulators for FWV-39 and FWV-40 maintain >27 psig for at least one hour when isolated from their air supply.

I i

s 3/4 7-5 Amendment No. 77,38 l CRYSTAL RIVER - UNIT 3

PLANT SYSTEMS CONDENSATE STORAGE TANK

~

LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be O'PERABLE with a minimum contained volume of 150,000 gallons of water.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:.

With the condensate storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

Restore the CST to OPERABLE status or be in HOT SHUTDOLN a.

within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or b.

Demonstrate the OPERABILITY of the condenser hotwell as a backup supply to the emergency feedwater system and restore the condensate storage tank to OPERABLE. status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

~.

~

SURVEILLANCE REOUIREMENTS 4.7.1.3.1. The condensate storage tank shall be demonstrated OPERABLE at least or.ce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume to be within its limits when the tank is the supply source for the emergency feedwater pumps.

The condenser hotwell shall be demonstrated OPERABLE at least

4. 7.1. 3. 2 once per 12 hcurs by verifying a, minimum contained volume of 150,000 gallons of water whenever the condenser hotwell is the supply source for-the emergency feedwater system.

CRYSTAL RIVER - UNIT 3 3/4 7-6 '.

a i

! 3/4.4 REACTOR COOLANT SYSTEM 1

BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolan.t loops in op-eration, and maintain DNBR above 1.30 during all normal operations and i

anticipated transients. With one reactor coolant pump not in operation in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based i

on RCS Flow and AXIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERMAL POWER for the num-ber of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive.

1 A single reactor coolant loop provides sufficient heat removal capabili-ty for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a DHR loop into operation in the shutdown ecoling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3/4.4.2 RELIEF VALVES - SHUTDOWN The pressurizer code safety, 'ves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig.

Each safety valve is desir;ned to relieve 317,973 lbs per hour of saturated steam at the valta's setpoint.

The relief cepacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating DHR loop, connected to l

the RCS, provides overpressure relief capability and will prevent RCS l

overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS f"a being pressurized above its safety limit of 2750 psig.

The c.4 ained relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.

I Demonstration of the safety valves' lift settings will occur only during l

shutdown and will be performed in accordance with the provisions of Sec-tion XI of the ASME Boiler and Pressure Code.

3/4.4.3 RELIEF VALVES - OPERATING The power operated relief valve (PORV) operates to relieve RCS pressure below the setting of the pressurizer code safety valv's.

This relief e

4 l

valve has a remotely operated block valve to provide a positive shutoff l

l CRYSTAL RIVER - UNIT 3 8 3/4 4-1 Amendment No. J7, 38 i

REACTOR COOLANT SYSTEM BASES capability should the PORY become inoperable.

The electrical power for both the relief valve and the block valve is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage. path.

~

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydrau-11cally solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.

The low level limit is based on providing enough water volume to prevent a pressurizer low level or a reactor coolant system low pressure condi-tion that would actuate the Reactor Protection System or the Engineered Safety Feature Actuation System as a result of a reactor scram.

The high level limit is based on maximum reactor coolant inventory assumed in the safety analysis.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients. Operation of the power oper-ated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam gen-erator tubes is based on a modification of Regulatory Guide 1.83, Re-vision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degra-dation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of : characterizing the nature and cause of any tube de-gradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these cheniistry limits, 1

localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limi-tation of steam generator tube leakage between the primary coolant CRYSTAL RIVER - UNIT 3 8 3/4 4-2 Amendment No. 38

.' d i T.* n m

.f 1) $, k REACTORC00LbihSY$7bj I

8 BASES

- system and the seconda ry coolant system (prima ry-to-second a ry leak-age = 1 GPM).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to wi thstand the loads imposed during nomal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant.

Leakage in excess of this limit will require plant shutdown and an un-scheduled inspection, during which the leaking tubes will be located and pl ugged.

Operational experience has shown nct tube defects can be the result cf unicue c:erating cencitions and/or physical arrancements in specific li:.ited areas of :ne steam genera: Ors (fer examcle, tubes adiacent c :ne ic::en inspec: ten lane or tubes wncse lhn tube su:por: plate hole is nc:

icrcached bu: crilled).- A full inspection cf all cf :ne tunes in such spe-

' cific limited areas will provide complete assurance that degraced or cefec-

-ive tuces in these areas are detectec.

Because no creci is taken for

nese cis-inctive tubes in the constitution cf tne first sample er its re-suhs, :ne recuirements for :ne first sample are unenengec.

This require-ment is essentially equivalent to anc mee:s the inten cf :ne requirements se: for:n in Regulatory Guice 1.53, " Inservice Inspection cf Pressuri:ed na:er Reac:cr Steam Generator Tubes", Rev.1, July 1975, anc does not re-et.ce :ne r.argin of safety providec Dy those requirements.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube exami-nations.

Plugging will be required for all tubes with imperfections ex-ceeding the plugging limit of 40% of the tube nominal well thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Wheaaver the results of any steam generator tubing inservice inspec-tion fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, labora-tory examinations, tests, additional eddy-current inspection, and revi-sion of the Technical Specifications, if necessary.

The steam generator water level limits are consistent with the initial conditions assumptions in the FSAR.

CRYSTAL RIVER - UNIT 3 B 3/4 4-3 Amendment No. 33, 38

I REACTOR COOLANT SYSTEM t

BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detectior, systems required by this specification are provided tc detect and monitor leakage from the Reactor Coolant Pressure These detection systems are consistent with the recomendations Scunda ry.

of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems". May 1972.

l t

3/4.4.6.2 OPERATIONAL LEAKAGE FRE55URE BO'JNDARY LEAKAGE of any magnitude is unacceptable since it may bd indicative.of an impen:ing gress failure of the pressure boundary.

Tnerefore, the presence of any PRESSURE BOUliDARY LEAKAGE requires the j

l l'

unit tc. be promptly placed ir. COLD SHUTD0h'!i.

I l

Industry experience has showr, that, whi'.e a limited amount of leakage f

1l i

be is expected from the RCS, the UNIDENTIFIED LEAKAGE portion of th s can This threshold value is

~

reduced to a threshold value of less than 1 GPM.

I sufficiently low to ensure early cetection cf aeditional leakage.

I The total steam generater tute leasage ".icit of 1 GPM for all steam generators ensures that tne c: sage contrit::.: tion from tube leakage will t,e lir.ited to a small fracticn of :a-t 100 Timits in the event cf either l

The 1 GPM limit is a steam generator tube ru;ture or steam line break.

l consister. with the assumptiens used in the analysis of these accidents.

I The 10 GPM,10ENTIFIED LEAKAGE limitation provides allowance for a 1;mited amount of leakage fr m kr.own sources whose presence will not ir.terfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limit cf 10 GPM restricts operation with a seals in excess of 10 GPM.

l

.cici RCS leakage from all R: pun:

I B'3/4 4-4 CRYSTAL RIVER - UNIT 3 1

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P00R ORIGINAL. p.. i I.

Amendment No. 5. M e 14 l

6-3

-....___...-._.___.________.__.__.-__'i!T3_ _. _. _. _ _ _ _ _ _, _ - _ _.. -.... _.. _ _, _ _ _ _.... _ _. -. _ _ _ _

CRYSTAL RIVER U.

TABLE 6.2-1 MINIM)M SHIFT CREW COMPOSITION #,,

APPLICABLE M00ES LICENSE CATEGORY 1, 2, 3, & 4 5&6 SQL 2

1*

OL 2

1 Non-Licensed 3

1 Operations Tech. Advisor 1

0

~

• Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling Indi-vidual supervising CORE ALTERATIONS after the initial fuel loading.

1. Shift crew composition may be less than the minumum re-quirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on duty shift crew members provided immediate action is taken to re-store the shift crew composition to within the minimum requirements of Table 6.2-1.

CRYSTAb RIVER - UNITI 3C ! ': o

,6-4,

Amendment No. 38 l

• ?'N ! U l t; 5 iht)h)

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the mini-mum qualifications of ANSI N18.1-1971 for comparable positions, except for the Chemistry and Radiation Protection Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the Operations Technical Advisor, who shall have a Bachelor's degree, or the equivalent, in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Plant Mana-ger and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI NI8.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Nuclear Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, except for Fire Bri-gade training. sessions which shall be held at least quarterly.

6.5 REVIEW AN3 AUDIT 6.5.1 PLANT REVIEW COMMITTEE (PRC)

FUNCTION 6.5.1.1 The Plant Review Committee shall function to advise the Nuclear Plant Manager on all matters related to nuclear safety, i

COMPOSITION 6.5.1.2 The Plant Review Committee shall be composed of tie:

i Chairman:

Technical Services Superintendent j

Member:

Operations Superintendent Member:

Technical Support Engineer l

Member:

Maintenance Superintendent Memoer:

Chemistry and Radiation Protection Engineer Member:

At large (Designated by Chairman)

ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PRC Chairman to serve on a temporary basis; no more than two alternates shall participate as voting members in PRC activities at any one time.

MEETING FREQUENCY 6.5.1.4 The PRC shall meet at least once per calendar month and as con-l vened by the PRC Chairman er his designated alternate.

CRYSTAL RIVER - UNIT 3 6-5 Arendment No. E. U, H, 38 i

L

i s

1 ADMINISTRATIVE CONTROLS OUORUM 6.5.1.5 A quorum of the PRC shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The Plant Review Committee shall be responsbile for:

r' Review of 1) all procedures required by Specification 6.8 and a.

changes thereto, 2) any other proposed procedures or changes thereto as determined by the Nuclear Plant P.anager to affect nuclear safety.

b.

Review of all proposed tests and experiments that affect nuclear Safety.

Review of all proposed changes to the Appendix "A" Technical c.

Specifications.

3 i

t d.

Review of all preposed changes or modificatiens*to plant systems or equipment that affect nuclear safety.

Investigation of ali violations of the Technical Specifications e.

including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Manager, Nuclear Operations and to the Chairman of the Nuclear General Review Committee.

f.

Review of events requiring 24-hour written notification to the I

Commission, Review of facility operations to detect potential nuclear g.

safety hazards.

l Performance of special reviews, investigaticns or analyses and h.

reports thereon as requested by the Chairman of the Nuclear General Review Committee.

i.

Feview of the Plant Security Plan and implerenting procedures and shall submit recommended changes to the Chairman of the Nuclear General Rev.ew Committee.

Review of the Emergency Plan and implementir.g procedures and J.

shall submit recomm. ended changes to the Chairman of the Nuclear General Review Committee.

6-6 Amendment No. E,2 5 a CP.YST*L RIVER - UNIT 3

I I

' ADMINISTRATIVE CONTROLS l

RECORDS 6.5.2.11 Records of NGRC acti"'

!s shall be prepared, approved and distributed as indicated below, a.

Minutes of dach NRCC meeting shall be prepared, approved and forwarded to the Senior Vice President Engineering and Con-struction within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 3.5.2.8 above, shall be prepared, approved and forwarded to the Senior Vice President-Engineering and Construction within 14 days following completion of the review.

Audit reports encompassed by Section 6.5.2.9 above, shall be c.

forwarded to the Senior Vice President Engineering and Construc-tion and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION

+

6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a.

The C'ommission shall be notified and/or a report submitted pursuant to the requirements of Spedification 6.9.

b.

Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Comission shall be reviewed by the PRC and submitted to the NGRC and the Manager, Nuclear Operations.

l l

l 6-11 Amendment No. E,2E

,; CRYSTAL RIVER - UNIT 3

o ADMINISTRATIVE CONTROLS 6'7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The facility shall be placed in at least HOT STANDBY within one hour.

b.

The Safety i.imit violation shall be reported to the Corsnission, j

the Manager, Nuclear Operations and to the NGRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the PRC.

This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures i

and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the NGRC and the Manager, Nuclear Operations within 14 days of the violdion.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and main-tained covering the activities referenced below:

The applicable procedures recommended in Appendix "A" of

~

a.

Regulatory Guide 1.33, November,1972.

b.

Refueling operations.

Surveillance and test activities of safety related equipment.

c.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

9 Systems Integrity Program implementation.

h.

lodine Monitoring Program implementation.

i 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PRC and approved by the Nuclear Plant Manager prior to implementation and reviested periodically as set forth in administrative procedures.

CRYSTAL RIVER - UNIT 3 6-12 Amendment No. 5 L3, 25,38

-