ML18046B408

Safety Evaluation Supporting Amend 15 to License DPR-67. Calvert Cliffs Sample Tech Specs Re Reactor Protection Sys & Engineered Safety Features Actuation Sys Encl

ML18046B408
Person / Time
Site:	Calvert Cliffs, Saint Lucie, Fort Calhoun
Issue date:	06/28/1977
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML18046B409	List: ML18046B408 ML20003G040 ML20003G047
References
TAC-7089, NUDOCS 8104280143
Download: ML18046B408 (18)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSlor.:

ENCLClSURE 2 WASHINGTON, D. C. 20565 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 15 TO LICENSE NO. DPR-67 FLORIDA PO\\.JER & LIGHT COMPANY ST. LUCIE PLANT UNIT NO. l DOCKET NO. 50-335 INTRODUCTION By applicatiQn dated April 20, 1976, Florida Power & Light Company (FPL) requested amendment to the St. Lucie Plant Unit No. l license.

The amend-ment would modify the Technical Specifications to.allow any one of the four Reactor Protection System (RPS) channels and one of the four Engineered Safety Feature Actuation System (ESFAS) channels *to be in the bypassed con-dition indefinitely. The current specification requires that after one hour the inoperable channel must be in the tripped condition.

DISCUSSION As the result of the February 18, 1976meeting, the NRC staff stated that.

the Technical Specifications dealing with the RPS and ESFAS would require

. that an inoperable channel be placed in the tripped condition within one hour: Thus; Technical Specifications for St. Lucie Unit No. l were issued on March 1, 1976, with the issuance of the facility license. The bases for the staff position were:

a.

The one hour requirement is consistent with the specifications imposed on other pressurized water reactor nuclear steam system suppliers; and

b. The staff reviewed the plant as a two-out-of-four system, not as a two-out-of-three system with an ins ta 11 ed spare.

18~ l 0._41 2 8 f[) \\'"\\ '>.>

\\)O~ -

r*

.e The staff agreed to re-open the review if FPL provided infonnation sufficient to show that the RPS and ESFAS is a two-out-of-three system with a spare. Also upon completion of the review, if the staff found that the RPS and ESFAS was an acceptable two-out-of-three system with a spare, the Technical Specifications would be appropriately modified.

In addition, the staff stated that in order for the system to be an acceptable two-out-of-three system with a spare, the licensee*

should demonstrate that with any one chanr.el in the bypassed co-ndition, no single failure could cause the RPS or ESFAS to fail to perform its protective function when required.

. The staff, in accordance with the previous commitment, re-opened the review of the Technical Specific~tion limitation after receiving the April 20, 1976 FPL letter. Basically, the issue to be resolved may be restated as follows:

To take credit for a two-out-of-three logic with an installed spare on the RPS and ESFAS, the licensee must demonstrate that the system-satisfies the single failure requirements of IEEE Std 279-1971.

EVALUATION In August 1976, a member of the NRC staff conducted an onsite review of the actual physical separation f"eatures of the RPS.and ESFAS at the St. Lucie facility to determine conformance with the licensee's stated criteria and with the reauirements of IEEE. Std 279-1971.

The onsite review included a detailed verification of the location and physical separation of sensors, instrument lines, transmitters, and

• electric cables routed through conduit and covered cable trays within containment for the following safety-related sensors:

l. Pressurizer Pressure Sensors;

2.

Resistance Temperature Detectors;

3.

Steam Generator Pressure Sensors;

4.

Steam Generator Differential Pressure Sensors;

5.

Ex-core Neutron Detectors; and

6.

Steam Generator Level Sensors.

The physical independence of containment electrical penetrations and -*

the routing of the four 11 independent 11 safety-related channels through the cable spreading room to the control room was also reviewed in detaiL In addition, the staff reviewed the routing and location of RPS and ESFAS instrumentation panels and the electric cables observing their potential for damage from *postulated breaks in high energy *pipe lines and from other hazards.

r -

e e Based on the onsite visit and the NRC staff review of licensee's sulxnittal of April 20, 1976, the following considerations required evaluatio.n:

1.

Th~ licensee indicates that a fourth channel of protective instrumentation was installed as an optional spare at an appreciable cost in order to give the utility added flexi-:

bility of opefations.

• Yet no other Combustion Engineering

{CE) design has a three chan~el RPS and ESFAS.

In addition, at St. Lucie cables for the four channels are routed in two groups.

For example, cables in channels MA *and MC are grouped together and likewise, MB and MD are grouped 'together.

The greatest separation distance is maintained between the two groups.

Four equally spaced, independent cable groups would have provided greater four channel physical separation and electrical independence.

2.

Within containment, a possibility exists that a main steam line break could disable two redundant safety-related steam generator differential pressure transmitters at the 62-foot elevation.

In addition, a high energy line break near piping penetrations #1, 2, 3, and 4 could possibly disable redundant RPS and ESFAS cables. Al.though the transmitter and cable failures may cause the respective channels to fail in the 11 safe 11 direction, the staff has not acceptei:f a 11fail.. safe 11 design as a design basis. Therefore, high energy line haiards may affect the minimum acceptable redundancy required by IEEE Std 279-1971.

3.

The RPS panels at.St. Lucie are practically identical to those provided for the Calvert Cliffs Nuclear Plants of Baltimore Gas and Electric Company.

A review of these panels was con.-

ducted at Calvert Cliffs and the results are applicable to St. Lucie.

The wiring terminated at the RPS panel allows an associated circuit to be routed with tw9 protection channel groups.

The wiring separation design criteria within the RPS cabinet can allow an associated circuit to be routed with another protection channel.

As a result of the existing separation within the RPS panel at Calvert Cliffs, the staff could not conclude that adequate four channel separation was maintained in the RPS cabinets. Because the associated circuits were not distinctively coded and because of inaccessibility to portions of the RPS panels, the staff could not verify these findings at St. Lucie. Therefore, the conclusions of the Calvert Cliffs review were applied to St. Lucie.

.. *4.

The RPS and ESFAS channel cables from the tran~mittcrs inside containment to the control room are routed in conduit or totally enclosed cable trays.

The cable trays and conduit were distinc-tively marked in most areas.

However, flammastic coatings of cables prevented precise cable tray identification.

In addition,.

since the cables could not be individually identified and traced, the actual routing could not be verified, except at the termina-tions. The licensee provided quality assurance certification showing that the cables were indeed designed to be installed in their respectively marked trays.

The staff audited a sample of cable trays for confirmation of the certification. The color code scheme of the cable trays coincided with the description on the quality control sheet.

However., at one location at the base of containment there appeared to be an error in the color code scheme of pressurizer pressure channels. Nevertheless,

• since the cables themselves had not been 11electronically traced 11 the ~taff could not be ~onclusivelv assured that t~~ cables were routed in their appropriate trays.

We consider that adequate cable separation is maintained for a two~out-of-four system; however,. this cannot be physically verified for a two-out-of-three system with installed spare.

5.

During the licensing review for the Calvert Cliffs plant, a potential single failure occurrence (hot short)"was discovered that would cause the four scram actuation 14elays (Kl, K2, K3, K4), a one-of-.six logic matrix, to de-energize {the four relays are in parallel). A change was made which split the logic..

matrix in half and powered Kl and K2 from one vital supply and K3 and K4 from a redundant supply. This change corrected the

• CE design; however, it was later determined by CE that a reassignment of logic power supplies located in different bays of the Reactor Protection System Cabin~t was necessary to avoid a spurious reactor trip. As a result of this change, possible damage in one bay of the RPS may be propagated to a redundant bay and/or impact two V*ital power inver:-ters.

Therefore, a potential exists for a single failure event to impact two channels.

6.

A review of the design of the St. Lucie Emergency DC Power system indicates that two emergency DC buses provide power to four inverters, which in turn supply power to the four RPS and ESFAS channels. Since this design does not provide

... complete independence to all four chilnnels, an overvoltage condition on one DC bus may be communicated to two channels through their respective inverters and damage ESFAS and RPS instrumentation circuits.

7.

One of the associated circuits in the control and safety-related consoles *Of the control room have instrument cables designated 11 !

11 These low 1 evel signal cables are fed by current 1 imited power supplies.

The licensee indicates that 11 IA 11 cables, once in the safety-related cable tray system; would be assigned only to the 11 IA 11 safety-related channel.

The 11 IA 11 cables should never be routed with any other safety-related cable.

However, the staff noted that when the 11 I 11 cables left their respective channels in the control room, no obvious separation existed.

For instance, 11 IA 11 and 11 IB 11 cables were brought close together in instrument panels.

The staff noted that a power supply failure could conceivably compromise redundant associated and possible safety-related instrument cables.

8.

The staff considers that a minimum of three ex-core nuclear power detector channels may oe required to detect certain transients and accidents, such.as rod ejection accidents, to prevent unacceptable core damage.

Indefinitely bypassing one of the detector channels has not been considered in combination with bther single failures.* Therefore, the staff would not allow an indefinite bypass of this parameter.

9.

Although the licensee has not requested that the Containment Spray Actuation System and the Recirculation Actuation System*

be considered for indefinite three-channel operation, four channels of these systems are necessary to.. satisfy the single failure crit'erion *

. Rased on the considerations noted above allowinri iridefinite bypass of one of the fout RPS or ESFAS is not justified. However, the St. Lucie four channel system of safety-relaterl instruMentation rloes have qreater inde-pendence than m~n.v other two-out-of-four systems and in itself justifies a reasonable outaqe allowance for testinq or maintenance of one channel.

Other1-1ise, required frequent testinq with the reactor at power cou1d re-sult in undesirable inadvertent reactor trips. The staff review conducted on these instrumentation systems for the Calvert Cliffs plants and _the

,,,:* St. Lucie plant has been essentially identical.

Our review concluded that bypass of one of the four chunnels may be permitted for test and maintennnr.e purposes for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Also, a channel muy be placed in trip for an indefi-nite period of time and while this condition exists, one additional channel may be placed in bypass for a period not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the sole pur-pose of performing tests and maintenance on that channel.

In addition, the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> bypass period has been approved recently for Calvert Cliffs Units No. l and No. 2, plants of a similar design, after a similar NRC staff review.

The operability of the RPS and ESFAS instrumentation and modi-fied bypass feature, as approved with this technical specification change, will assure that (1). the ~PS and FSF~S trtps wtll occur when the Monitored para-meter exceeds its setpoint, (2) the 2/3 or 2/4 coincidence loqic is maintained, *

(3) sufficient redundancy is maintained to permit a channel to be out of ser-vice (bypassed or tripped) for testing or maintenance, and (4) sufficient sys-tem functional capability is available from the diverse parameters.

ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change *in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4),

that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of the amendment.

CONCLUSION We have concluded, based on the considerations discussed above, that: (1) be-

. cause the amendment does not involve a signficant increase in the probability or consequences of accidents previously considered and does not involve a sig-

• nificant decrease in a safety margin, the amendment does not involve a signifi-cant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: June 28 1977

e ENCLOSURE 3 3/4.3. INSTRUMENTATION 3/4. 3. l REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.l.l As a m1n1mum, the reactor protective 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.

SURVEILLANCE REQUIREMENTS 4.3.l.l.l Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECr:, CHAtrnEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.l.2 The logic for the *bypasses shall be.demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days.

The total bypass function shall be demonstrated OPERABLE at least once per 18 ~onths during CHANNEL CALIBRATION testing of each

.channel affected by bypass operation.

4.3. 1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor 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 once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the 11Total No. of Channels" column of Table

3. 3-1.

CALVERT CLIFFS 3/4 3-1

~

-

--w+-<...,.-.*

~

I

T/\\BU J.3-1 n

RE/\\CTOR PROTECTlVI. JNSTRUMENT/\\TION

)>

r <

fT1 MINIMUM

o

-t TOTAL NO.

CHANNELS CHANNELS

/\\PPL I CABLE n

FUNCTIONAL UNIT OF Cll/\\NNELS TO TRIP OPERAOLE MODES ACTION r

01 01

1. Manual* Reactor Trip 2

2

. 1, 2 and

• l lll

2.

Power Level - High 4

2 3(f) l, 2*

2#

3.

Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1 ' 2 (e) 2#

e

4. Pressurizer Pressure - lligh 4

2 3

1 ' 2 2#

5.

Containment P.ressure - High 4

2 3

l ' 2 2#

6.

Steam Generator Pressure - Low 4/SG

• 2( b )/SG 3/SG 1, 2

'2n

7.

Steam Generator Water w

levP.1 - Low 4/Sf.

2/Sfi 3/SG l

• 2 2#

.J'>

w

8. Axial flux Offset 4

' ( r.)

.I 211 I

N

9. :I.

nwrm.il ;.1drgin/low Prc:;:;urc

'1 i ( d)

I, i:' ( e) 211 j

b.

Steam Generator Pressure Difference - lligh 4

2(a) 3 1 ' 2 (e) 2#

10.

loss of Turbine--Hydraulic Fluid Pressure - Low 4

2(c) 3 1

2#

n Tl\\BLE 3. 3-1 (Continu~

)>

• ~

REACTOR PROTECTIVE INSTRUMENTATION IT1

~

-I n

r MINIMUM TOTAL NO.

CHANNELS

• CHANNELS APPLICABLE VJ FUNCTIONAL UNIT OF CHl\\NNELS TO TRIP OPERABLE MODES

'ACTIO~

11. Wide Range Logarithmic Neutron Flux Monitor

a. Startup and Operating--Rate 2'9 of Change of Power - High 4

2(d}

3(f) l, 2 and *

b.

Shutdown 4

0 2

3, 4, 5 3

w

~

12. Reactor Protection System 6

1 6

1, 2*

4 w

Logic Matrices I w

13. Reactor Protection System 4/Matrix 3/Matrix 4/Matrix l, 2*

4 logic Matrix Relays

14. Reactor Trip Breakers 8

6 8

l, 2*

4

TABLE 3.3-1 (Continued)

TABLE NOTATION

• With the protective system trip breakers in the closed position and the CEA drive system ~apable of CEA withdrawal.

1. The provisions of Specifi~ation 3.0.4 are not applicable.

(a) Trip may b~ bypassed below lo""4 of RATED THERMAL Po~4R; bypass shall be automatically removed when THERMAL POWER is > 10 of RATED THERMAL POWER. * \\

(b)

Trip may be manually bypassed below 685 psia; bypass shall be automatically removed at or above 685 psia.

(c)

Trip may be bypassed below 15~; of*RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15~ of RATED THERMAL POWER.

(d)

Trip may be bypassed below 10-4% and above 12~; of RATED THERMAL POWER.

(e) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.

(f} There shall be at least two decades of overlap between the Wide Range Logarithmic Neut~on Fl~x Monitoring Channels and the Power Range ~eutron Flux Monitoring Channels.

ACTION 1 ACTION 2 ACTION STATEMENTS With the number of channels OPER~BLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OP~RABLE 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 HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

and/or open the protective system trip breakers.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable chanhel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then* be either restored to OPERABLE status or placed in the tripped condition.

CALVERT CLIFFS 3/4 3-4

~* ~...

I I

ACTION 3

.A.CTION 4 I

CALVERT CLIFFS TABLE 3.3~1 (Continued)

ACTION STATEMENTS

e.

b.

Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped-, as applicable) as that required by a. above for the inoperable channel.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

With the number of channels OPERABLE one less thatfl required by the Minimum Channels OPERABLE requirement, ver*ify compli-ance with the SHUTDOWN MARGIN requirements of Spe~cification 3.1.l.l or 3.1.1.2, as applicable, within l 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 channels OPERABLE one less tha=n required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may ;be

• bypassed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for surveillance testin:g per Specification 4.3.1.1.

3/4 3-5

INSTRUMENTf.TION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION

. 3.3.2. l The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column o1 Tabl~ 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY:

As. shown in Table 3.3~3~

ACTIGN:

a.

With an ESFAS instrumentation channel trip setpoint less 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 set-point 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 4.3.2.1.l Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the perfonnance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL. FUNCTIONA~ TEST operations during the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL 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.

4.3.2.l.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the 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 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.

CALVERT CLIFFS 3/4 3-10

.1

("")

TABLE 3. 3-3.

):>>

I <

n'1 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

o

--l

(""')

MINIMUM I......

TOTAL NO.

CHANNELS CHANNELS APPLICABLE e

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE

.MODES ACTION Vl

1.

SAFETY INJECTION (SIAS)

a.

Manual (Trip Buttons) 2 2

1 ' 2, 3, 4 6

b. Containment Pressure -

High 4

2*

3 l ' 2, 3 7*

c. Pressurizer Pressure -

Low 4

2 3

.1 ' 2, 3(a) 7*

2.

CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) 2 1

2 1 ' 2, 3, 4 6

w b

• Containment Pressure --

.i:-

High 4

2 3

1 ' 2' 3 11 w

I __,

3.

CONTAINMENT ISOLATION (CIS) a.. Manual CIS (Trip Buttons) 2 1

2 1, 2, 3. 4 6

b.

Containment Pressure -

High 4

2 3

1, 2. 3 7*

~

' ""Y:.

TABLE 3.3-3 (Continued) n ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

):>

r <

IT1

o

~

MINIMUM n

TOTAL NO.

CHANNELS CHANNELS APPLICABLE

-r FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION "Tl "Tl Vl

4.

MAIN STEAM LINE ISOLATION

a.

Manual (MSIV Hand Switches

• and Feed Head Isolation Hand Switches) 1/valve l/valve l/valve l, 2, 3, 4 6

b.

Steam Generator 4/steam 2/steam 3/steam l, 2, 3(c)

]ir Pressure - Low generator generator gene:--ator

5.

CONTAINMENT SUMP w

RECIRCULATION (RAS) a. Manual RAS (Trip

.i:::.

Buttons) 2 1

2 l ' 2, 3, 4 6

w

• I......

N

b.

Refueling Wat.er Tank - Low 4

2 3

1, 2, 3 7*

e

n TABLE 3.3-3 {Continued}

)>

I <

ENGINEERED SAFETY FEATURE ACTU/\\TION. SVSTEM lNSTRUMENTATIOU rr1

~

n MltllMUM r

t--4 TOTAL NO_.

CHANNELS CllANNELS APPLICABLE "Tl "11 fUNCT IONAL UN IT OF CllANNELS*

TO TRIP OPERABLE MODES ACTION V1

6.

CONTAINMEtlT PURGE VALVES ISOLATION a-.

• Manual (Purge Valve Control Switches) 2/Penetration l/Penetration 2/Penetration. 1 2, 3, 4 6

b.

Cont~inment Radiation -

lligh Area Monitor 4

• 2 3

.-6 8

w

7.

LOSS OF PO\\olER

-I

a. 4.16 kv Emergency Bus w

Undervoltage (Loss of I

Voltage) 4/Bus 2/Bus

. 3/llus 1 t 2 I 3 7*

w

b. 4.16 kv Emergency nus Undervoltage (Degraded 4/0us Voltage}

2/Bus 3/Bus

. l

• 2. 3 7*

j

I n

).:o r <

('Tl

• ~

n I t-t.,,.,,

U1 w -

.p.

w I __.

~

T/\\OLE 3.3-3 (Continued)

ENGHIEERED SAFETY FE/\\TURE /\\CTUATION SYSTEM INSTRUMEflT/\\TIOtl FUN CT ION/\\L UN IT B.

eves ISOLATION

a.

Manual (eves Isolation Valve Control Switches)

b.

Hcst Penetration Room/letdown lleat Fxchanger Room Pressure - lligh TOT/\\l NO.

OF CHANNELS l/ValVe Cl!/\\NMELS TO TRIP 1/\\falve 2

t1I NI MUM CllAtlNELS OPER/\\OLE

• l/Valve

. 3 J\\PPLI C/\\BL E MODES l, 2, 3,4 l,2,3,4 ACTION 6

1

~.*.,

.. *.. J.

TABLE 3.3-3 (Continued)

TABLE NOTATION (a)

Trip functt6n may be bypassed in this MODE when pressurizer pressure is < 1700 psia; bypass shall be automatically removed when pr_essurizer pressure is.:. 1700 psia.

(c) Trip functio~ may be bypassed in this MODE below 685 psia~ bypass*

shall be automatically removed at or above 685 psia.

• The provisions of Specification 3,0.4 are not a~plicable.

ACTION 6 -

ACTION 7 -

CALVERT CLIFFS ACTION STATEMENTS 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 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With the number of OPERABLE channels one less than the Total Number of Channels, operation maY ~roceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within l hour.

For the purposes of testing and maintenaoce, the inoperable channel may be bypa~sed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the i noperab 1 e channel shall then be either restored to OPERABLE st~tus or placed in the tripped condition.

b.

Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped. as applicable) as that required by a. above for the inoperable channel.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the ot_her inoperable challnel is pl~ced in the tripped coridition.

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ACTION 8 -

ACTION 11 CALVERT CLIFFS TABLE 3.3-3 (Continued)

With less than the M1nimum Channels OPERABLE, operation may continue provide the containment purge valves are maintained closed.

With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Mini~um Channels -OPERABLE requirement is demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; 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.Z.l.

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